ML20126B710

Exam Rept 50-266/OL-92-02 Administered on 920917 & 28-1002. Exam Results:Four SRO & Three RO Candidates Passed Written & Operational Exams

ML20126B710
Person / Time
Site:	Point Beach
Issue date:	12/04/1992
From:	Burdick T, Hansen J, Clyde Osterholtz, Reidinger T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20126B679	List: ML20126B677 ML20126B710
References
50-266-OL-92-02, 50-266-OL-92-2, NUDOCS 9212220183
Download: ML20126B710 (16)

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'U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-266/OL-92-02 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: Milwaukee, WI 53201 Examination Conducted: September 17 and Septamber 28 through October 2, 1992 RIII Examiners: /MM f I N J. L. Hansen Date r k ikJd.

es (C . C. Osterholtz Date m n/a Chief Examiner: 8) M..

• I L/ H h ?-

me. T . Reidinger Date Approved'By: An#2. L 2.-[ Q h L -

T. M. Burdick, Chief Date Operating-Licensing Section 2 Division of Reactor Safety Examination-Summang Examination administered on Scotember 17 and'Seotember 28 throuch October 2. 1992 (Report No. 50-266/OL-92-02): TheLwritten-examinations were administered on September 17,.1992 atLthe1 Inn On Maritime Bay, Manitowoc, WI. The operational exams were

< administered at the roint Beach Facility September 28 through October.2, 1992,.. U t r: examination was administered to two:-instant

- senior reactor ope' u tr.(SRO-I) candidates, two upgrade senior reactor operator.(Sun-U) candidates and six reactor operator (RO) candidates. I n a d d i t i o n ,: a retake simulator examination was administered to one-SRO-I candidate.

Results: The' retake;SRO-I candidate 1 failed. Four SRO and threo

RO candidates _ passed the written.and_ operational examinations.

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Two reactor operator candidates failed both'the written and operational examinations. One reactor operator candidate failed.

only.the written examination.

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PDR ADOCK 05000266_

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The following are examples of the strengths and weaknesses identified by the NRC evaluators.

Strenath

• The use of alarm resp nse books in determining plant status.

Weaknesses

• Some shift supervisot 'SRO) d nonstrated a weakness in proper usage of the notes and cautions contained in emergency procedures.

• Communications between crew members during dynamic simulator examinations was poor.

• The senior reactor operators' command and control abilities were deficient in that it lacked oversight and coordination and inhibited efficient use of all crew members.

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REPORT DETAILS

1. Examiners 1

J. Hansen, NRC C. Osterholtz, NRC R. Pugh, PNL

• T. Reidinger, NRC

• Chief Examiner l l

2. Exit Meetina l i

An exit meeting was held on October 2, 1992, with the i facility management and training staff representatives, to discuss the examiners' observations.

FRC representatives in attendance were:

J. L. Hansen, Examiner i C. C. Osterholtz, Examiner R. Pugh, Examiner 1 T. Reidinger, Examiner K. Jury, Senior Resident Inspector Licensee representatives in attendance were:

G. Maxfield, Manager-PBNP-F. Hennessy, Simulator Coordinator J. Becka, Manager-Regulatory Services R. Sizert, Manager-Training T. Vandenbosch-Training Coordinator The licensee representatives acknowledged the examiner observations discussed in Section 3 of this report.

3. Examiner Observations

a. Examination Development The majority of reference material submitted for examination development was considered adequate but with minor exceptions. Some examples are:

.' Technical Specifications had overy other page omitted in addition to the index.

Section 4 of the PBNP Administrative Procedures was not sent to the contract examiner as requested initially.

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• The set of logic prints went through sheet 24' while the set being used in the simulator only went through sheet 22 (AMSAC logic was missing).

• The facility Initial Operator Test Questions bank was of limited uso due to the majority of questions not being in the objective-format, having inadequate refereT;cs, and not identifying specific knowledge and LW ilty tasks.

• The users guide for the simulator malfunction book was omitted.

Although the licensee's traini,' staff provided the NRC with good support and cooperati6n during the examination materials validation and examination administration, the following weaknesses were noted:

• The amount of time required to validate scenarios ,

during preparation week and to reinitialize the I simulator between scenarios during exam week was unusually long due to a continued simulator operator unfamiliarity with simulator operations.

• A training department instructor potentially compromised examination security with two days of classroom instruction to candidates after signing

, the exam security agreement. The training  ;

department identified the potential. compromise to l the NRC. The NRC determined that no compromise '

existed after reviewing the training schedule and classroom material,

b. Operatino Examination Adm!mistration (1) During-operating examination admin'istration, the following strengths and weaknesses were noted:

Strenoths

• The use of alarm response books in determining plant status and taking appropriate actions.

• The SRO's use of health physics personnel 1to-monitor steam lines for radiation during accident' conditions.

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Egoknesses e The communication between crew members regarding plant and. equipment status during major transients.

e General knowledge of administrative requirementu including radiation protection and temporary modifications. A majority of Reactor Operator candidates admitted'a lack of training in adminit. rative procedures.

• Lack of overall crew coordination by the SRO's during major transients.

Several candidates commented that they could have used additional simul.ator time and better classroom instructions prior to the examination. They also indicated they were not fully aware of the NRC exam process.

(2) Materials provided for the candidates hse during examinations were not consistent with those

, available to operators in the control room. For example:

• Operator aids in the control room had not been placed in the simulator (i.e., operating instructions for the radiation monitoring console, dyno tape identifying an accumulator fill valve).

• P & ID's used by the candidates for JPM questions were not in the same order as those in the control room (several candidates had trouble-locating the auxiliary feedwater print).

• One controlled copy of technical specifications used by the candidates had pages reversed and out.of order.

Several procedural problems were noted and 1 identified to the training personnel. For example:

• A revision to Critical Safety Procedure (CSP)

S.1, " Response to Nuclear Power _

Generation /ATWS," permitted a transition to 1

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complete the immediate actions of emergency operation procedure ( EO'P) 0, " Reactor-Trip or-Safety Injection," raises two concerns:

a. The caution was not consistent with the Westinghouse Owners. Group (WOG)

Functional Response (FR) Guidelines.

The caution in FR-S.1, " Response to Nuclear Power. Generation - ATWS,"' states.

"If an SI signal axists'or occurs, the immediate actions of E-0, " Reactor Trip or Safety Injection," should be performed while continuing with.this guideline." The WOG FRG background for FR-S.1 states "This step (the caution) should be performed in parallel-with the subsequent steps of this guideline as manpower and time permits." The facility elected to direct the control room personnel to complete the immediate-actions of EOP-0, Reactor Trip or Safety Injection, and then immediately. continue with the CSP-S.1 procedure. The justification given for the deviation was potential operator confusion.in performing parallel procedures.- '

b. The revision to CSP-S.1 was effective on September 30, 1992, but several licensed.  !

training personnel were not aware of the change prior to implementation which resulted in no training being provided to licensed operators and candidates concerning this change.

• The control room use of I&C Routine Maintenance Procedure ICP-10.2, " Procedure for Removal of a Safeguards or-Protection Sensor from Service." While this is titled as-an I&C routine maintenance procedure, it is actually.used asian abnormal operating procedure.by-operations.- The fact that a note existed'within the procedure to reduce power to less than'80,%:if certain conditions' occur due to instrument failures further.

emphasizes this concern. This power.

reduction requirement'was not cross '

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referenced or identified in other operations-instructions (i.e., Alarm Response Books, Abnormal Operating Procedures).

• Sork classification guidance in emergency-plan implementing procedure EPIP 1.2, " Event classification," appeared inadequate (i.e., a' loss of all main feedwater and auxiliary.

feedwater for greater than i hour results in declaration of a General Emergency). No guidance was given for this situation for less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in duration. This could result in support (i.e., the Technical Support Center) not being activated until a-General Emergency is declared. This apparent lack of guidance also occurred in other classification areas which were identified to the licensee.

• There are significant differences (up to 8 deg. F) from verious subcooling margin indicators available to the operator with no procedural guidance as to which should be used to comply with the EOP requirements.

4. Written Examination-Administration

, The post examinLtion review of the written examination by the NRC identified the following deficiencies in the candidates' knowledge. A majority of the candidates failed to provide the correct response for cach particular knowledge area identified. This information is being provided as input to the licensee's system approacn to training (SAT) process. No response is required.

RO weaknesses identified:

• The response of the Rod Control System while

, recovering a dropped rod.- (RO question 014)

• The individual who must grant permission for waiver of independent verification requirements due to unacceptable radiation exposures.

(RO question 18)

• The administrative requirements of-exceeding Technical Specification oxygen levels during plant startup. (RO question 22) 2

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• The reactor coolant pump (RCP) function as_a leakage-barrier when the pump motor is uncoupled from the shaft. (RO question 29)-

• The immediate actions necessary to mitigate a main feed pump recirculation valve failure while at i 100% power. (RO question 33)

• The Technical Specification actions required to be taken during refueling operations with no operable source range detectors.- (RO question 34)

• The requirement for stopping RCP's prior to depressurization during specific accident conditions and the basis for that requirement.

(RO question 37)

• The administrative requirements concerning Temporary Modifications and recirculation sump "B" concerns at power. (RO question 50)

• The response of the overtemperature Delta T trip setpoint to changes in plant conditions and i equipment failures. (RO question 53) i

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• The amount of time required for decay heat l generation to decrease below 1%'following.a reactor trip from 100% power, steady. state conditions. (RO question 61)

• The administrative requirements for "Short Term ,

Relief" as defined in PBNP 4.4, " Main Control Room Conduct and Access." (RO question 60) l

• The immediate actions required with two rod bottom-lights lit, control bank "D": fully withdrawn and a rod out demand signal while the reuctor is at 60%

power. (RO question 97)

• The functions of the Incore-Instrumentation System. (RO question 98)-

SRO weaknesses identified:

• The Reactor Coolant Pump (RCP) function as a leakage. barrier when-the pump motor is uncoupled from the shaft.- (SRO question 29)

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• The requirement for stopping RCP's prior to

, depressurization during specific accident conditions and the basis for that requirement.

. (SRO question-37) 3.

• The Auxiliary Feedwater System response to Lo-Lo steam generator levels while shutdown with a simulated level signal to "B" steam generator.

(SRO question 45) e The administrative requirements for "Short Term Relief" as defined in PBNP 4.4, " Main Control Room Conduct and Access." (SRO question 68)

• The actions required during recovery from a Loss ,

of Offsite Power. (SRO question 82) >

e The Hydrogen Monitor default calibration parameters following a loss of power. (SRO question 83) e The use of AOP-9C, " Degraded RHR System Capability," Attachment B, " Time After Shutdown' L Curve," to determine the tima to reach saturation

! following a loss of RHR at Mid-Loop. (SRO question 85)

• The activation criteria for the Emergency Operations Facility. (SRO question 100)

5. Written Examination Review Licensee representatives reviewed the written examination prior to administration with appropriate changes being incorporated into the examinations by the NRC at that time. Following administration of the-written examination, the facility was given a copy of the RO and SRO examinations and answer keys for review.

The facility's post examination comments and the NRC resolutions are contained in Enclosure 2 of this report.

6. General Observations The following observations were mada by'the examiners-while administering examinations:-

• Security and Health Physics personnel were courteous and cooperative-in assuring minimum delays when touring the plant.

• Control room personnel provided good support during candidate performance of inplant JPM's.

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Enclosure 2' Facilftv Comments and NRC Resolution of Comments RO/SRO Ouestion 50:

In accordanca'with PBNP 3.1.10, " Temporary Modifications," which of the following is used to address sump "B" recirculation concerns at power?

a. The temporary modification tag is posted on the EL 66' personnel hatch outside door. '

b. The temporary modification is identified in a separate section of the Night Order Book.

c. The Modification Engineer is required to track these temporary modifications and provide a list to the (DSS needs to be defined) monthly.

d. The temporary modification work control paperwork is maintained by the responsible engineer.

ANSWER:

i a.

REFERENCE:

PBNP 3.1.10, Rev. 13 p.8 para.4.6.3 KA 194001K102 3.7/4.1 EHliP COMMENT / RECOMMENDATION: >

There should be two correct answers, "A" and "B"._ The wording of this question is misleading. .It leads the examinee to believe

-there is a problem with Sump B recirculation phase that requires

i. a Temporary Modification. Based on PBNP 3.1.10, Rev. 13, page 2,

! paragraph 2.8, "A summary of the temporary 1 modification describing its important physical and operating characteristics l-- shall'be made on a night order book form, approved by the DSS, I

and placedfin the night order book."

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NRC RESOLUTION:

i, Comment accepted. Choice "A" or "B".will be accepted as correct answers.

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k 4 RO/SRO OUESTION-li:

Which of the following initially indicates that the PRT rupture disk has ruptured following a pressurizer PORV failing OPEN?

a. PRT temperature decreasing.

b. Relief line temperature decreasing.

c. PRT level low.

d. Pressurizer level decreasing.

EEB:

a.

REFERENCE:

TRHB 10.3 Rev. 1, p.7 Steam tables KA 000008A108 3.8/3.8 PBNP COMMENT / RECOMMENDATION:

There should be two correct answers, "A" and "B". When the PRT rupture disc blows, the temperature in the PRT and the relief line should decrease, because both lines will feel the effect of the constant enthalpy throttling process. Since.the relief line and the-PRT will be at approximately the same-pressure, their temperatures will be approximately the same.

' NRC RESOLUTION:

Comment accepted. Choice "A" or "B" will be accepted as correct answers.

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7 RO OUESTION 90:

The following conditions exist:

G-01 diesel Loaded to 2810KW G-01 diesel DC Power Failure Alarm (alaru circuit-breaker is open and will not reset)

Which of the following describes how the loss of DC power would affect diesel generator operation.

a. The diesel would continue to run but would stop in response to an overspeed condition.

b. The diesel would stop due to low fuel pressure from the DC fuel oil pump.

c. The diesel can be stopped by manually moving the fuel rack to stop fuel injection.

d. The diesel can be stopped by placing the mode selector switch to " Exercise".

ANSWEE:

c.

REFERENCE:

OP-11A Rev. 21 P.4 LP0133 Rev. 10, pg.76 KA 064000K104 3.6/3.9 PBNP COMMENT / RECOMMENDATION:

There should be two correct answers, "A" and "C". The overspeed trip is mechanical and not dependent on DC control power. Since no_ governor control power is available, an overspeed condition may occur.

NRC RESOLUTION:

Comment accepted. Choice "A" or "C" will be accepted as correct answers.

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RO OUESTION 98:

Which of the following is accurate with regard to either the incore thermocouples or the incore flux mapping system?-

a. Uses incore thermocouples to measure the temperature of coolant-flow axially through the core.

b. Verify QPTR limits, but will not determine the position of misaligned control rods,

c. Monitors fuel assembly outlet temperaturos,

d. Used to determine core thermal power.

ANSWER:

C.

REFERENCE:

TRHB 13.2 Rev. 1, p.1 KA 017020K102 3.3/3.5 PBNP COMMENT / RECOMMENDATION: [

Eliminate question no. 98. Strict interpratation of the question has no answer that applies to both.

NRC RESOLUTION:

Comment is rejected. Choice "C" is the correct answer. The question as written with the "either/or" statement does not indicate that an answer must apply to both the incore thermocouple and incore flux monitoring' systems to be correct.

This wording of the stem was recommended by the facility during the examination pre-review.

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J Enclosure 4 1 SIMULATION FACILITY REPCRT Facility Licensee: Point Beach Nuclear Plant Facility Licensee Docket Nos. S0-266; 50-301 R

Operating Tests Administered On: Week of Septenber 28, 1992 I While conducting the simulator portion of the operating tests, the following items were observed and identified to the facility:

ITEM DESCRIPTION Initial Condition (IC) Several IC sets listed in the PBNP Sets simulator Book and identified for use by the NRC examinerfin ',

scenarios could not be used as written. The simulator instructor said the IC sets were old and not-updated (i.e., IC-7, IC- 16) . Only-one IC was available for plant end-of-life (EOL) conditions.

Team Program Pailure The program suggested by the computer engineers to record and printout selected primary and secondary parameters during-scenario execution-failed to' work-properly and created a simulator i~

crash prior to completing a scenario. The program could not be used to retain information following the failure.and there was-no other program available to accomplish the request.

S / G Tube Leak'Malf. This malfunction was not available for use as it has not been -

validated since upgraded by-Westinghouse. ~ This lengthened.the scenario validation process due to having to adjust a S/G Tube Rupture malfunction to the correct leakage setpoint, c

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Rod' Control System The rod control system caused rods .

to move automatically even when.the 1 Rod Select Switch was in manual 1or selected to individual rod groups.

This occurred-during scenarios and JPM's.

Core Exit Thermocouples The core exit thermocouples (CETs) failed from approximately 1400 degrees F. to approximately 200 degraea F. during.two scenarios.

This forced early terminatico of those scenarios due to the CET's being needed-for response to the- 1 Emergency Operating Procedures.

I Reactor Makeup System The reactor makeup system did not automatical~ ' stop diluting _when the reactor sakeup batch integrator ,

reached the . ero setting. Tho' i operator ~was required'to manually take the makeup control switch-to l the stop position to secure the dilution. This occurred several times.

Annunciator Alarms The annunciator panel alarms in the plant differ from the alarm in the simulator. The alarm in the simulator does not differentiate between panels as do the contro!

room alarms. This was a hinderance to the candidatec as they responded.

to alarms during the scenarios.

SI/CTMT Iso. Reset The reset buttons-for Safety Injection _and ContainmentLIsolation were not working properly. ._Roset could not be accomplished during.a scenario. Between scenarios,-the buttons were repaired Nr I&C and from that point on ap; eed to work correctly.

Plant Process _ Computer The. Simulator Process Computer failed several times, delay _ing__the-validation and examination process.

Also, the-simulator process computer does not-have a--" screen print" available which is_available_

'in the control room. ,

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Plant-Cycles Cycle 16 is being used_in the simulator while the plant is at cycle 20.

CCW Tlow Indication During_ Scenario NRC-CC-1A (10), the CCW-flow-indication began oscillating =after the reactor trip.-

The oscillations were between 1500 and 2000 gpm early in the scenario and increased to between 0 and 2000 gpm just prior to termination of the scenario.

Loose Parts Monitoring The noise generating portion of the loose parts monitoring system was -

not modeled ir cgards to audible monitoring.

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