ML17310A829
| ML17310A829 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 11/04/1993 |
| From: | Morrill P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17310A828 | List: |
| References | |
| 50-528-OL-93-03, 50-528-OL-93-3, NUDOCS 9312030086 | |
| Download: ML17310A829 (31) | |
Text
ENCLOSURE 1
Examination Report No.: 50-528/OL-93-03 Facility:
Palo Verde Nuclear Generating Station, Units 1, 2, and 3
Facility Licensee:
Facility Docket Nos.:
Arizona Public Service Company P.O.
Box 53999, Station 9082
- Phoenix, Arizona 85072-3999 50-528/529/530 Facility License Nos.:
NPF-41/51/74 Twelve initial licensed reactor operator and five senior reactor operator upgrade examinations were administered at the Palo Verde Nuclear Generating
- Station, Units 1, 2, and 3, on September 13 through 21,
- 1993, near Wintersburg, Arizona.
In addition, a senior operator limited to fuel handling retake operating examination was administered on September 21, 1993.
Examiners:
Approved by:
Philip H rill,'hief Operatio s Section David B. Pereira, Chief Operator Licensing Examiner John Kramer, Operator Licensing Examiner Philip HorrIll, Chief, Operations Section Tom Vehec, PNL Operator Licensing Examiner J'ckolaus, PNL Operator Licensing Examiner Date
~Summar:
Examinations on Se tember 13 throu h Se tember 21 1993.
Re ort No. 50-
~50 5 Written examinations and operating tests were administered to five senior reactor operator upgrade (SROU) and twelve reactor operator (RO) applicants.
Two out of five SROUs and nine out of twelve ROs passed these examinations.
In addition, a retake operating examination was administered to a senior operator limited to fuel handling (LSRO).
The LSRO passed his examination.
The licensee's simulator support staff provided excellent technical staff support during the scenario development and during the operating examination administration.
The facility did not complete a thorough review of the written examination during the preparation week.
After the written review was completed the facility had additional concerns.
As a consequence, the Chief Examiner resolved several comments on the written examination during the week after the 93i2030086 93ii05 PDR ADOCK 05000528 PDR
t 4
examination review week.
Based on the practice at other facilities, it appeared that the presence of a licensee supervisor during the written examination review would have minimized the lack of communications.
The license examiners identified two training concerns during the examination administration.
The first concern was that the Emergency Operations Procedure did not state which electrical feeder breakers could be tripped to cause a
scram following an anticipated transient without scram (ATWS).
The second concern was that during one scenario administered to three groups of candidates, none of the candidates observed that instrument failures had occurred which should have caused a reactor trip.
The examination team determined that the facility should consider (I) clarifying which feeder breakers should be tripped following an,ATWS condition, and (2) conduct additional training to enable candidates to deal with multiple instrument failures and to improve diagnosis of the cause of safety system actuations.
Safet Si nificant Issues None.--
j
REPORT DETAILS Examiners D. Pereira, Chief Examiner, RV J.
Kr amer, Licensing Examiner, RV T. Vehec, Licensing Examiner, PNL J. Nickolaus, Licensing Examiner, PNL P. Horrill, Chief, Operations
- Section, RV Persons Attendin the Exit Meetin on Se tember 22 1993 NRC:
D.
J.
J.
- Pereira, Chief Examiner Kramer, Licensing Examiner
- Sloan, Senior Resident Inspector Palo Verde:
J.
E.
R.
F.
p.
J.
J'.
D.
A.
B.
B.
R.
R.
L.
p.
D.
W.
, R.
J.
F.
H. Levine, Vice President Nuclear Production Firth, General
- Manager, Nuclear Training
- Nunez, Manager, Operations Training Riedel, Unit I Manager, Operations Wiley, Unit 2 Manager, Operations
- Dennis, Manager, Operation Standards Scott, Unit 3 Assistant
- Manager, Operations Gouge, Director, Plant'Support
- Peroutka, Supervisor, Training Grabo, Supervisor, Nuclear Regulatory Affairs Picchiottino, Supervisor, Simulator Support Bouquot, Supervisor, guality Assurance Audits
- Fountain, Supervisor, guality Assurance Honitoring Speight, Shift Supervisor, Unit 2 Operations Coffin, Engineer, Nuclear Regulatory Affairs Martin, Senior Nuclear Instructor, Training Department Hontefour, Coordinator, Owners Services Henry, Site Representative, Salt River Project Draper, Site Representative, Southern California 'Edison Gowers, Site Representative, El Paso Electric
3.
Initial Written Examinations a ~
Examination Administration The licensee reviewed the RO and SRO initial written examinations at the Palo Verde Nuclear Generating Station Units 1, 2, and 3 on August 30 and 31, 1993.
The licensee review team consisted of a senior nuclear instructor, a simulator support instructor, and a
unit shift supervisor.
During the written review and comment
- period, no licensee training supervisor was present to resolve licensee validity questions.
Subsequently, the Chief Examiner resolved additional licensee comments on written questions during the week of September 6,
1993, after licensee training supervision became involved.
Even though the late resolution of written comments had no effect on preparing or presenting the written examinations, the examiners thought that facility written test review and comments should have been resolved during the preparation week.
b.
To prevent a reoccurrence in the future, the Chief Examiner suggested that additional time be taken by the facility for licensee review and comment of the written examination.
In addition, the Chief Examiner suggested that the licensee have a training supervisor present to review questions for validity and to assure that licensee concerns were adequately resolved.
The written initial examinations were administered on September 13, 1993 to five SRO and twelve RO candidates at the site.
At the conclusion of the written examinations, the facility training staff were given copies of the written examinations as administered, and the candidates'xaminations.
The licensee presented written comments which are addressed in Enclosure 3 of this report.
Examination Results Four out of five SRO candidates passed the written examinations.
All twelve RO candidates passed their written examinations.
One SRO candidate failed his written examination.
C.
Generic Weaknesses The NRC conducted a post-examination comparison of written test results and determined that there were several areas of knowledge that appeared weak.
In particular:
(1) 33% of the operators did not know the automatic actions of the plant following an auxiliary feedwater actuation signal (AFAS); (2) 88% of the operators did not identify a condition that requires a trip of a reactor coolant pump; (3) 59% of the operators did not know that the unit 2 atmospheric dump valve controller at the remote shutdown panel will return to the control room (CR) position after restoration of electrical power to the control,ler; (4) 65% of the operators did not know the
I
st'arting criteria for a RCP based on previous start attempts; (5) 71X of the operators did not know that loads would not automatically sequence on Unit 1 bus 1-E-PBA-S03 when it is being energized from a unit 3 diesel generator during a station blackout because the 4.16 KV breakers had their 86 lockout relays tripped; and (6) 71X of the operators did not know that a failed level transmitter (LT-227) in the chemical and volume control system would cause the-divert valve to divert, but not affect automatic makeup.
In the administrative portions of the written examination there was one area of weakness.
Only 20X of the SROs correctly knew their escort responsibility when escorting visitors in protected/vital areas.
Conversely, 92X of the ROs knew their escort responsibilities.
The Chief Examiner concluded that the training program coverage of these, areas had not been uniformly effective.
4.
0 eratin Examinations a
~
Scenario Examinations Administration Scenario examinations were reviewed with the licensee during the week of August 30 through September 2,
- 1993, by the use of the Palo Verde simulator.
The scenario events and transients were validated by the NRC examiners to ensure that operator responses were correctly identified.
This, NRC validation of scenarios was assisted by a facil,ity senior simulator operator',
an operations, shift supervisor, and a senior training 'instructor.
Scenario examinations were administered on September 14 and 15, 1993 to the five SROs and twelve ROs.
Two out of five SROs and eleven out of twelve ROs candidates passed this portion of the operating examination.
The NRC examination team identified two concerns during the administration of the simulator scenarios.
The examiners also observed that some candidates had difficulty diagnosing events and following procedures.
The first concern was that the Emergency Operations Procedure did not state which electrical feeder breakers should be tripped to de-energize the control element drive motor (CEDM) motor-generator (MG) sets following an anticipated transient without scram (ATWS).
During a scenario in which an ATWS occurred, the operators tripped the reactor by removing power to one 13.8 KV bus.
The two 13.8 KV busses supply power to the reactor coolant pumps.
The operators used procedure 41EP-IE001, Emergency Operations, in which the secondary operator safety function flowchart (SFFC) required the operator to de-energize load centers L03 and L10 for at least 5
- seconds, or manually open the reactor trip breakers.
In this
- scenario, simulator malfunctions failed both the main feeder breaker
I C
I I
to L10 and the feeder breaker above L10 in the closed position.
The Emergency Operations Technical Guideline (EOTG),
40DP-9AP06, describes alternative methods to shut down the reactor if the CEAs fail to automatically trip.
The EOTG statement for de-energizing the MG sets states:
"Remove the power from the MG sets by opening the L03 and L10 supply breakers on the control board."
All three crews given this scenario completed the following actions.
The secondary operators attempted to de-energize the MG sets by first attempting to open L03 supply breaker (which opened) and L10 supply breaker (which was failed shut),
then the feeder breaker to L10 (which also failed shut).
All three secondary operators decided to de-energize the 13.8 KV electrical bus feeding L10 by opening the bus supply breaker (S02).
When the supply breaker
- opened, power to the L10 load center was lost.
None of the crews sent an operator to manually open the reactor trip breakers, which was-one of the expected success paths established during the preparation week.
When questioned, the candidates stated that they would open breakers as far as necessary to de-energize the L10 load center.
The Emergency Operations Procedure (41EP-1E001) is not specific in describing which breakers to trip to de-energize load centers L03 and L10, nor does the facility training material describe any limitations on how to achieve the unloading of load centers L03 and L10.
The examiners concluded that the licensee should ensure that clear and consistent steps for removing power from the MG sets are stated in procedures and used for training.
Additional guidance in the training lectures should be given to ensure that alternative methods of shutting down the reactor (such as emergency boration or dispatching an operator to the MG sets) are examined and considered after the removal of power steps are exhausted.
The second major concern was observed during the same scenario.
None of the candidates observed that instrument failures had occurred which should have caused a reactor trip.
This scenario contained two simultaneous instrument failures followed by a Loss of Coolant Accident (LOCA) and an ATWS.
In this scenario, a loss of power to the 120 VAC bus PNA-D25 occurred which de-energized channel A of the plant protection system (PPS).
At the same time, a ¹1 steam generator (SG) wide range level channel D failed low.
Because of the loss of power to bus PNA-D25 and the coincidence failure of channel D wide range level, a two out of four logic was satisfied for a PPS low level steam generator
¹1 reactor trip and an Auxiliary Feedwater Actuation Signal (AFAS).
After five to seven
- minutes, a
During the preparation week, facility representatives agreed with the examiners that a manual scram was the expected response due to the instrument failures.
Subsequently, all three crews failed to identify that the instrument failures should have caused a reactor trip.
No candidates recommended a.
reactor trip nor a Technical Specification shutdown.
One crew realized that an AFAS had occurred and correctly diagnosed the cause of the AFAS.
They observed that a two out of four logic was completed for low level in SG number I; This crew recognized the loss of channel D wide range level in SG number I combined with the Channel A PPS failure caused the AFAS signal.
They completed satisfactory mitigation strategy for the AFAS condition.
- However, this crew did not identify or acknowledge that the instrument failure should have also caused a scram.
Followup questioning by the NRC examiners determined that this crew concluded that 'shutdown within one hour due to entry into a Technical Specification 3.0.3 was the correct response.
In summary, one crew out of three realized that a two out of four coincidence had occurred for the SG low level logic.
Only one crew realized that an AFAS had occurred and performed satisfactory mitigation.
This crew indicated knowledge of the plant conditions and correctly determined that the plant process parameters were not in jeopardy.
An apparent weakness is indicated in training on instrument failures associated with PPS failures.
In regards to candidates having difficulty diagnosing events -and following procedures, the examiners observed several examples of poor performance.
In one scenario a steam line break inside containment was diagnosed as an uncontrolled steam release to the atmosphere.
In another'cenario a candidate deactivated an AFAS by placing the pump switches in pull-to-lock without using the procedure.
That action made the auxiliary feedwater system unavailable for any subsequent AFAS.
During a
SG tube leak scenario, the candidates did not appear to recognize that almost all auxiliary feedwater flow was going to the lower pressure.
SG and that the low pressure SG flow needed to be throttled to obtain flow to the higher pressure SG.
This was significant because the high pressure SG had the tube leak and atmospheric dump valves were used to lower that SG pressure.
Injection of cold auxiliary feedwater would have made an atmospheric release unnecessary.
In a faulted'and ruptured SG scenario one candidate deliberately departed from the emergency operating procedures.
This action led to an unnecessary re-pressurization of the reactor coolant system.
The examiners concluded that several candidates had difficulty using the emergency operating procedures, did not consistently use procedures when appropriate, or had difficulty comprehending the overall plant response.
b.
Walk Throu h Examinations Administration Walk through examination review with the licensee was conducted
0
during the week of August 30 through September 2,. 1993.
Minor changes were made during the simulator review of the Job Performance Measures (JPHs) by the examiners.
JPM examinations were'dministered on September 16 through 21, 1993 to the five SROs, twelve
- ROs, and a Senior Operator Limited to Fuel Handling (LSRO).
One JPH, No.
PNL004, entitled "Parallel the "8" motor generator (HG) set",
using 400P-9SF03, "CEDH HG Sets Operations" was difficult for at least one candidate.
This candidate performed the JPH on the wrong HG set,, that is, the "A" MG set in spite of the examiner providing cues to the contrary.
The labeling of the CEDM HG set control panels appeared to have contributed to this incorrect performance.
Each individual CEDM HG set control panel had a nameplate which describes the "A" or "B" HG set.
This nameplate was difficult to read, and due to the fact that the "A" HG set.remote control panel was in the center of the "B" HG set control panel makes a possible identity error very likely.
Subsequently, the licensee informed the Chief Examiner that an event occurred in Unit 2 where the wrong HG set had been secured in 1992, which caused a reactor trip.
The licensee stated that corrective action to make the label plates clearer was in progress.
During the NRC post-examination review the examiners found that one system/JPH evaluation question was invalid.
The question related to JPM PNL005, Respond to Loss of Shutdown Cooling (SDC) Flow.
The question asked why the RCS temperature must be kept below 200 F when the containment spray (CS) pump is aligned to SDC.
The answer, "to ensure that discharge piping integrity and operability requirements are met prior to entering mode 4," was not consistently supported by Technical Specifications or licensee procedures.
The Technical Specifications state that mode 4 is entered at 210 F, not 200 F.
The containment spray (CS) system Technical Specification requires that the CS system be operable in mode 4, unless it is being used for shutdown cooling.
Consequently, by Technical Specifications, the CS pump could be used for SDC above 200 F.
Procedure 410P-1ZZOl, Cold Shutdown to Hot Standby Mode 5 to Mode 3, does have a
precaution to maintain RCS temperature below 200 F when a
CS pump is used for SDC, but does not explain why this precaution exists.
Other licensee procedures state that the CS system must be cooled below 200 F when venting the system which implies that operation above 200 F is allowable.
The facility staff stated that the basis for the requirement in 410P-12ZOl is to avoid exceeding the engineering design basis temperature for the CS system piping.
The examiners concluded that this level of knowledge is not appropriate for operator testing.
Based on Technical Specifications the question has no correct answer.
Based on plant procedures, a'correct answer rests on
knowing the engineering design temperature for the CS system which is not an expected knowledge for an operator.
For this reason this question was deleted from the examination.
During performance of several
- JPHs, the examiners observed that some candidates had difficulty following procedural
- steps, and demonstrated a lack of knowledge of system operation.
One candidate failed to enter the correctly calculated CPC Aximuthal Tilt allowance value in a critical step of procedure 72ST-IRX03.
In a different step of the same JPM, the candidate attempted to bypass both Plant Protection System (PPS) channels A and B.
PPS circuitry does not allow more than one channel to be in bypass at a time.
In a different JPH, the candidate failed to return the "System Mode Selection" switch to "Operate."
This was a facility identified critical step.
During administration of several other walk through tests, other JPH failures were observed.
The candidate(s) did not restore an emergency core.'cooling system to operable status following a
- LOCA, did not recognize out of limit Technical Specifications for Azimuthal Tilt, and failed to trip reactor coolant pumps during a
control room evacuation.
These JPH failures appeared to the examiners to be due to inattention to details in the procedures, and lack of familiarity with plant equipment or 'operation.
Several system questions were failed due to an apparent lack. of system knowledge.
All five SROs, nine out of twelve ROs, and the Senior Operator Limited to Fuel Handling passed their JPH walk through examinations.
5.
Exit Meetin An exit meeting was held by the NRC examiners with representatives of the licensee's staff on September 22, 1993 to discuss. the NRC examination.
The licensee did not identify as proprietary any of the materials provided to or reviewed by the examiners during the examination.
e ENCLOSURE 3
FACILITY COMMENTS/NRC RESOLUTION OF FACILITY COMMENTS 50-528/OL-93-03
ENCLOSURE 3 Page 1 of.6 FACILITY COMMENTS/NRC RESOLUTION OF FACILITY COMMENTS RO Examination uestion:
14 SRO Examination uestion:
16 Facility Comment:
Our Training Program Objectives allows use of associated Tech-Spec LCO to determine LCO compliance.
No reference was provided.
We do not expect the operators to have the heatup and cooldown rates memorized.
They will keep the rate IAW procedural guidelines which they will have in hand while performing a heatup or cooldown.
Facility Recommendation:
questions are invalid.
NRC analysis:
Based on the Knowledge and Abilities Catalog, NUREG-1122, and 10 CFR 55.43, an SRO should recognize the maximum heatup and cooldown rates allowed by Technical Specifications.
The Knowledge and Abilities Catalog number 002000G005 states that "Knowledge of limiting conditions for operations and safety limits" has an importance of 3.6/4.1 for RO/SRO respectively.
In
- addition, KA 005000A101 states the ability to predict and/or monitor changes in parameters (heatup/cooldown rates) to prevent exceeding design limits associated with operating the shutdown cooling system controls has an importance of 3.5/3.6 for RO/SRO respectively.
10 CFR 55.43(b)(2) states that
. the written examination for a senior operator will include items from the facility operating limitations in the Technical Specifications and their bases.
Although important knowledge for the RO, the RO is supervised by SROs.
, On this basis, we conclude it is not essential that the RO know the heatup and cooldown rates.
NRC Resolution:
Retain the question and answer for the SROs.
Delete the question for the ROs.
RO Examination uestion:
29 SRO Examination uestion:
32 Facility Comment:
Our Training Program Objectives allows use of associated Tech Spec LCO to determine LCO compliance.
No reference was provided.
We do not require our operators to memorize the SDM vs Temperature graphs.
They are required to determine if SDM is adequate by comparing plant data to the graphs.
The operators refer to 72ST-IRX09 for determination of shutdown margin determination'.
Facility Recommendation:
guestions are invalid.
I I
ENCLOSURE 3 Page 2 of 6 FACILITY COMHENTS/NRC RESOLUTION OF FACILITY COHHENTS NRC analysis:
Knowledge of the reasons for initiating emergency boration has an importance of 4.1/4.4 for ROs and SROs respectively (KA 000024K301).
If shutdown margin requirements cannot be met at IOOX power an emergency boration is required.
This requirement is a Technical Specification LCO and should be recognized promptly.
All the distractors require knowledge of the action
~ statement for a limiting condition for operation associated with technical specifications.
Me conclude that the question requires a greater depth of knowledge of Technical Specifications than required of the RO and is therefore not appropriate for the RO.
- However, 10 CFR 55.43(b)(2) states that the written examination for a senior operator will include items from the facility operating limitations in the technical specifications and their bases.
NRC Resolution:
Retain the question and answer for the SROs.
Delete the question for the ROs.
RO Examination uestion:
34 SRO Examination uestion:
37 Facility Comment:
Consideration:
Insufficient data was provided to determine the Haximum power level.
In order to determine this, Hain Feedpump Suction pressure must be considered.
65X power is the DESIGN power level for a single Hain Feedpump.
70X is the HAXIHUH power level per EER 86-FT-021 IF suction pressure can be maintained
>300 psig.
Our Lesson Plan reflects BOTH values.
Facility Recommendation:
Accept Answers B and C.
NRC analysis:
During the preparation week, operations shift supervisors stated that based on prior operating experience there would be adequate main feedwater pump suction pressure with one main feedwater pump removed from service to achieve 70X power.
The stem of the question provides sufficient data to answer the question.
However, facility training material may have confused some of the operators because their lesson plan reflects that both 65X (answer "b") and 70K (answer "c") are correct answers.
4 NRC Resolution:
Agree with facility recommendation.
Change the answer key to accept "b" or "c" as the correct answer.
ENCLOSURE 3 Page 3 of 6 FACILITY COMMENTS/NRC RESOLUTION OF FACILITY COMMENTS RO Examination uestion:
57 SRO xamination uestion:
52 Facility Comments:
Our Training Program Objectives do not require this specific knowledge.
Facility Basis:
An Operator will always have a procedure in hand when starting an RCP(including during EOP's).
The procedure will provide the necessary guidance and prerequisite for the operator to determine when a pump start is allowed.
This also has a low K/A value (2.6/2.9).
Facility Recommendation:
guestions are invalid.
NRC analysis:
The starting of a RCP causes stresses on the motor windings thermally and mechanically.
Unnecessary starts should be avoided.
The ability to explain and apply all system limits and precautions (KA 003000G010 importance of 3.3/3.6 for RO/SRO respectively) will prevent equipment damage.
In addition the knowledge of RCP performance and design attributes (ie.
starting requirements) will ensure the RCP will not be damaged, (KA 003000K614 importance of 2.6/2.9 for RO/SRO respectively).
In accordance with the Examiners
- Handbook, NUREG/BR-0122, section 2.2.2.3, the KA should have a
rating of at least 2.5.
The question exceeds this requirement for both ROs and SROs.
When starting a
RCP the body of the procedure does not address the starting limits of the RCP.
The operators must rely upon their knowledge of the limitations and precautions to prevent pump or motor damage.
NRC Resolution:
Retain the question and answer as given.
RO Examination uestion:
65 SRO Examination uestion:
57 Facility Comment:
Our Training Program Objectives do not require this knowledge.
We expect the operator to go to the procedure and determine the requirements for check valve use in clearances.
Facility Recommendations:
guestions are invalid.
NRC analysis:
Plant-wide generic KA 194001K102 states that knowledge of tagging and clearance procedures has an importance of 3.7/4. 1 for RO/SRO respectively.
Senior reactor operators should be able to recognize what can be used as a boundary for a clearance (i.e. check valves, motor operated valves, or air operated valves that fail open) since they have the final authorization for the hanging or removal of tags for the unit area of responsibility.
10 CFR 55.43(b) states that the written examination for a senior operator will include items from administrative procedures for the facility.
Since the RO cannot be the Responsible Supervisor for the final authorization of the clearance, we conclude it is not essential the RO know when check valves can be used as a clearance boundary.
I
ENCLOSURE 3 Page 4 of 6 FACILITY COMMENTS/NRC RESOLUTION OF FACILITY COMMENTS NRC Resolution'.
Retain the question and answer for the SROs.
Delete the question for the ROs.
SRO Examination uestion:
59 No Facility Comment or Recommendation.
NRC analysis:
Three of the answers state a chemistry parameter is within a "specified limit:" Since the answers do not identify which chemistry limit
("steady state" or "transient") is being referred to, there are three possible correct answers.
NRC Resolution:
Delete the question.
RO Examination uestion:
68 SRO Examination uestion:
61 Facility Comment:
This question is outside the scope of an operators responsibility.
He has no job task to know what the OPS managers responsibility regarding Night Orders.
The operators task is that he should review the Night Order book after shift turnover.
Facility Recommendation:
guestions are invalid.
NRC analysis:
To ensure proper turnover of pertinent information, accurate and complete status of logs, night orders and other turnover items must be maintained.
However, it is not the responsibility of the operators to ensure the night orders are reviewed routinely by the operations manager.
NRC Resolution:
Agree with facility recommendation.
Delete the question for the ROs and SROs.
RO Examination uestion:
74 Facility Comment:
There is NO Cor'rect Answer.
What really happens in this condition is the Modulation Permissive is locked in.
SBCS continues to function normally.
Facility Recommendation:
guestion is invalid.
NRC analysis:
Due to an error, PT-1027 was used in the stem of the question instead of PT-1024.
Therefore, using the information provided in the
- question, there are no correct answers.
NRC Resolution:
Agree with facility recommendation.
Delete the question.
J
ENCLOSURE 3 Page 5 of 6 FACILITY COMMENTS/NRC RESOLUTION OF FACILITY COMMENTS RO Examination uestion:
82 SRO Examination uestion:
79 Facility Comment: Considerations:
With Condenser Vacuum 8 8.5 inches Hg and increasing, SBCS will have a
condenser interlock rendering SBCS valves 1001-1006 inoperable.
A turbine trip without SBCS operable may result in a Reactor Trip on high Pressurizer pressure due to inadequate heat removal.
Answer B may be considered acceptable based on Alarm Response for window 6A16B First Priority Action 5'l.
Answer D may be considered acceptable based on 41AO-lZZ07-Note preceding step 2.1.
Facility Recommendation:
guestions are invalid.
NRC analysis:
In the stem of the question, a turbine trip setpoint has been exceeded requiring a turbine trip, but the turbine has failed to trip.
The question calls for the FIRST action the operator should perform with a given set of conditions.
Based on the annunciator response for windows 6A16B (SBCS COND INTLK) and 6A16D (COND VAC LO) which both have a setpoint of approximately 5.0 inches HgA, an operator may trip the turbine (answer "a") or possibly take manual control of SGPV 1007 8
SGPV 1008 (answer "b").
NRC Resolution:
Change the answer key to accept "a" or "b" as the correct answers.
RO Examination uestion:
84 SRO Examination uestion:
81 Facility Comment: Considerations:
- 1. There are two different PZR pressure instruments on each channel (wide range and narrow range).
The stem of the question does not specify which instrument failed. If WR failed, then no automatic reactor trip would have been called for.
S/G level will rapidly decrease.
A reactor trip is Imminent.
- 3. If 2 instrument channels have failed, the reactor should have automatically tripped.
If it doesn',
you are in Tech Spec 3.0.3.
and a Unit Shutdown is required if not fixed in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
lj I[
0 I
ENCLOSURE 3 Page 6 of 6 FACILITY COHHENTS/NRC RESOLUTION OF FACILITY COHHENTS Facility Comment:
Continued:
Based on this, Answer A is a correct answer as the operator tries to stabilize feed flow to the S/G.
Answer C is also correct based upon the reactor may be approaching a trip setpoint due to the severe underfeed condition.
1 Facility Recommendation:
Accept answers A and C.
'RC analysis:
With the controlling channel on SG-1 failing high, the operator should take manual control of the steam generator level and attempt to recover level to prevent an automatic reactor trip.
In some cases, the steam generator may be in a severe underfeed condition and the operator would realize that manual control would be unsuccessful and a manual reactor trip would be appropriate.
NRC Resolution:
Agree with facility recommendation.
Change the answer key to accept "a" and "c" as the correct answers.
RO Examination uestion:
98 SRO Examination uestion:
93 Facility Comment:
Our Training Program Objectives do not require this specific knowledge.
Our Objectives allow use of reference material to determine stabilizing action.
Facility Recommendation:
guestions are invalid.
NRC analysis:
This question was based on the knowledge of parameters which require an immediate reactor coolant pump trip.
In accordance with 10 CFR 55.41(a) and 55.43(a),
a written examination question is not required to be linked to a training program objective.
The Knowledge and Abilities Catalog system generic KA 000015G011 states that the ability to -recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operation conditions has an importance of 3.5/3.6 for RO/SRO respectively.
The operators should be familiar enough with major pieces of equipment (such as reactor coolant pumps) to recognize failure or trip criteria.
NRC Resolution:
Retain the question and answer as given.
ENCLOSURE 4
SIMULATION FACILITY REPORT 50-528/OL-93-03
ENCLOSURE 4
SIMULATION FACILITY REPORT Facility:
F i'liy0 k
Palo Verde Nuclear Generatin Station Operating Tests Administered on:
Se tember 14 throu h 21 1993 I
This form is to be used only to report observations.
These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b).
These observations do not affect NRC certification or approval of the simulation facility other than to provide information that may be used in future evaluations.
No licensee action is required in response to these observations.
While conducting the simulator portion of the operating tests, the following items were observed:
ITEM DESCRIPTION 1.
Alarm printer failed to operate for 2 scenarios.
2.
The balance of plant console phone did not work.
.3.
A recorder failed on the Emergency Core Cooling System panel.
The simulator staff responded promptly to all of the above failures.