ML20148K297

Forwards Comments Concerning Written Exam Given to Plants Operators on 871214.Comments Divided According to Type of Test Administered W/Cross Ref to Redundant Questions to Assist Examiners Grading Exams

ML20148K297
Person / Time
Site:	Oconee
Issue date:	12/18/1987
From:	Tuckman M DUKE POWER CO.
To:	Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
Shared Package
ML15239A017	List: ML15239A016 ML15239A018 ML20148A019 ML20148K297
References
NUDOCS 8803310070
Download: ML20148K297 (20)

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e ENCLOSURE 3 Duios Powisit GOMPANY PRODUGTION SUPPORT DEPARTMENT OGONEE TRAINING GENTER r.o. nox 1430 MENEGA. H.C. 20070 December 18, 1987 Mr. John Munro operator Licensing Section Chief, Region II U.S. Nuclear Regulatory Commission 101 Marietta Street, Suite 3100 Atlanta, GA 30323 Attention:

Mr. William Dean Chief Examiner

SUBJECT:

NRC WRITTEN EXAM REVIEW Please find attached the comments concerning the Written Examination given to the Oconee operators on December 14, 1987.

These comments have been divided according to the type of test administered with cross references to redundant questions to assist the examiners grading the examinations.

Most comments are of a technical nature and provide additional information useable as additional correct answers.

References to the source (s) of this additional information have been provided.

It was noted that the examiners made a genuine effort to use the training material provided by the oconee Training Center and in s majority of cases referenced objectives used to train the Oconee operators.

The review, however, did identify several questions on each exam which seemingly goes beyond what an operator should be held responsible..4 of OMP 2-1 provides guidance on "Procedural Items Which All Licensed operators Shall Have Committed To Memory".

A list of questions which we feel do not apply using the guidance of OMP 2-1 has been attached.

A copy of each examination has been provided to the Duke Power Program Development Group for comments on exam construction and composition.

The comments are intended to be constructive and to assist in improving the NRC's Exam Bank as it is developed.

When received by the Oconee Training Center, the comments will be forwarded to you.

8803310070 880219 PDR ADOCK 05000269 V

PDR

m tir. John ' Munro NRC Written Exam R5vicW Pcg3 2 If there are any questions concerning these comments, please contact Larry Hindman at the Oconee Training Center, (803) 882-6150, Ext. 2252.

sincerely,

N.5.

A-'

' ' gL3 Mike S. Tuckman Station Manager Oconee Nuclear Station MST/MDT/src Attachments cc:

T. Barr L. Hindman File

[

ATTACHMEh? 1 COMMENTS REGARDING THE REACTOR OPERATOR LICENSING EXAMINATION Category 1.0 Principles of Nuclear Power Plant Operation, Thermodynamics, Heat Transfer, and Fluid Flow 1.18 a Another acceptable answer is "just critical".

A just critical reactor will behave as a subcritical reactor with regards to changes in neutron population change and, therefore, startup rate.

Both a subcritical reactor and a just critical reactor would asymptotically approach zero.

The answer key implies that a critical reactor would not be acceptable since the startup rate plot would have to asymptotically approach zero.

The plot provided in question 1.18 asymptotically approaches zero and no time reference is made.

Reference:

OP-OC-SPS-RT-SM, page 21 1.20 The answer key incorrectly references the CST (Condensate Storage Tank) as the suction source for EFW.

The normal suction source is the UST (Upper Surge Tank).

Reference:

OP-OC-SPS-SY-EF, page 14 1.21 Other acceptable answers would be a statement of the time dependence and power dependence / independence nature of these fission product poisons.

The answer key addresses the fission yield and absorption cross sections but fails to address the decay schemes involved in the production / removal of these poisons which in turn affects the time and power characteristics of each poison

Reference:

OP-OC-SPS-RT-FPP, (page 24 is summary) 1.23 The "black" APSRs originally installed at Oconee used silver, indium, and cadmium in their poison section - not just silver as indicated on the answer key.

Reference:

OP-OC-SPS-THF-PD, page 12 (nre1287.com 3)

ATTACC(ENT 1 Category 2.0 Plant Design Including Safety and Emergency Syst:ms 2.03 b The terminology, "KVI's", caused confusion.

When referring to KVIA, KVIB, KVIC, and KVID inverters, the term Vital (Bus)

Inverters is used.

Reference:

OP-OC-SPS-EL-VPS OP/1&2-3/A/1107/04 2.05 b&d Another acceptable combination of answers would be "(b) steam demand" with the turbine header pressure the "(d) same" as setpoint.

OP-OC-TA-NT states the OTSGs may become high level limited.

This is true for Units 2 and 3 because of the present operating levels in their OTSGs.

Unit 1, however, has undergone a steam generator chemical cleaning process which has substantially reduced the operating OTSG levels.

Unit 1 is nuw capable of undergoing the transient in question 2.05 without reaching the high level limit and producing the MWe error signal which modifies the turbine header pressure.

Units 2 and 3 are scheduled to undergo the chemical cleaning process.

Reference:

NSM-2113 (brief description attached) 2.08 This question was very eanfusing for the candidates (and the proctors).

Attempts at clarification led to further confusion as to what was being asked.

Some candidates were led to believe "loss of normal AC power" meant the main transformer was unavailable with the startup transformer available (therefore RCPs would continue running) while others were unsure whether the motor driven emergency feedwater pumps had power.

Realization that EFW pump run-out is a potential problem should be a sufficient answer for this question.

(The answer key states a verbatim answer is not required.)

Another possible answer would be OTSG overfill leading to carry-over and damage to the turbine driven emergency feedwater pump as a result of a loss of IA due to loss of AC power.

Reference:

OP-OC-SPS-SY-EF, page 68A l

OP-OC-TA-AT, page 22 2.13 The purification demineralizer resin is saturated with Lithium to l

aid in pH control of the Reactor Coolant.

Lithium-7 is preferred to Lithium-6 because of the difference in the amount of tritium (nrcl287.com 4)

ATTACHMENT 1

-production as stated in the answer key.

(Boron is correct)

Reference:

OP-OC-SPS-CH-PC, pages 11 and 14-15 OP-OC-CF-RC, page 9 L

2.14 The BWST will automatically be aligned to the HPI pump suction during ES actuation.

It is during this condition that the concern for LDST hydrogen overpressure exist.

A statement of "during ES actuation" is equivalent to addressing the "BWST gravity head".

Reference:

OP-OC-SPS-Sy-HPI, page 19 2.19 The term "alternate source" misled both RO candidates.

The question would be better worded using "alternate suction source".

One candidate asked for and received clarification enabling him to answer the question according to the key.

The other candidate was not made aware of this clarification and thus interpreted the question to imply that the "alternate source" was another unit's EFW.

This interpretation is based on actions taken in EP/1-2-3/A/1800/01.

With this interpretation he addressed the 6 foot minimum UST level and maximum flow of 500 gpm/ pump.

Reference:

EP/1-2-3/A/1800/01, Section 502 OP/1-2-3/A/1106/06, Encl. 3.4 2.21 This answer is incorrect.

While there is a line from the LPI system to provide pressurizer auxiliary spray, this line is not used because the LPI pumps cannot produce sufficient driving head to provide spray flow.

Instead HPI auxiliary spray flow is provided from a bypass on the injection nozzle warming lines.

(The lesson plan referenced by the answer key has been corrected.)

Reference:

OP/1-2-3/A/1103/10, Encl. 4.3 Category 3.0 Instruments and Controls 3.13 The ansvar hey for the Reactor Operator exam provides only valve numbers while the Senior Reactor Operator exam provides valve numbers and name/ description of the valves.

Either valve number or name/ description of the valves should be sufficient.

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(nrc1287.com 5) hs.

r ATTACHMENT 1 NOTE:

The name/ description for MS-126 and MS-129 on the SRO exam are incorrect.

They should read:

MS-126, Main Steam to Aux. Steam Control MS-129, Main Steam to Aux. Steam Control

Reference:

OP/1-2-3/A/1106/06, Limit & Precautions 3.14 Additional acceptable answers could include MS-93 failing open (loss of IA with loss of AC power) and the available DC Auxiliary Oil pump.

Although not specifically listed in OP-OC-SPS-SY-EF, these design features do provide for the AC power independence of the TDEFWP.

Reference:

OP-OC-SPS-SY-EF, page 55 3.19 The Core Saturation Margin Program has been changed.

The answer key is correct for reactor power less than 2%.

However, if reactor power is greater than 2% all available (47) operable incores are used.

The revised method for ca]culating the margin is in the OP-OC-IC-RCI lesson contained in the update package dated 10-5-87.

Reference:

OP-OC IC-RCI, page 37A (attached) 3.22 Clarification of this answer is necessary.

During normal operation with the "A" HPI pump running, the "B" HPI pump control switch will be in AUTO.

The "C" HPI pump control switch is a spring-return to neutral switch.

In these positions simply taking the switch to the OFF position following ES reset will secure the associated pump.

If the "B" HPI pump control switch had been in the OFF position I

when ES was reset then it would be necessary to take the switch from the OFF pcsition and back to OFF j'.i order to secure the "B"

HPI pump.

l

Reference:

OP-OC-SPS-SY-HPI, page 27 3.24 Rather than specify "3 of 4 nozzles" an acceptable alternative would be the ability to cross connect headers in order to achieve HPI flow in "both headers".

This will accomplish the required flow through 3 of 4 nozzles when the break is assumed in the fourth nozzle.

Reference:

EP/1-2-3/A/1800/01, Section 505 l

inrcl287.com 6) l

ATTACHMENT 1 Category 4.0 Procedures - Normal, Abnormal, Emergency and Radiological Control 4.17 a This question is not specific enough to attain the desired (answer key) response.

The term "Emergency Header" was misinterpreted.

When addressing EFW flow through the Emergency Header some candidates assumed this to mean the Normal-Emergency Header rather than the Emergency-Emergency Header.

The Normal-Emergency Header would be used if a failure of FDW-315/FDW-316 were to occur when EFW flow is required.

If the "Emergency Header" is interpreted to be the Emergency-Emergency Header then anytime EFW is actuated the Emergency Header would be used (flow through FDW-315 and FDW-316).

Therefore, an acceptable answer would be loss of MFPs or EFW actuation.

Reference:

OP/1-2-3/A/1106/06, Encl. 3.5 (step 2.2)

OP-OC-SPS-SY-EF 4.17 b This question is not specific enough to attain the desired (answer key) response.

Additional information is required to inform the candidates on plant conditions (i.e.)

MFW available ? /

EFW in auto ? / EFW in manual ?).

Due to the confusion from 4.17a, recognizing 4.17b is a continuation of that question is not sufficient.

Only after looking at the answer and reference on the key were the conditions clear.

Because of the reasons given an additional answer would be 240 inches on the XSUR.

This level would be maintained if the EFW system is automatically controlling OTSG level with no RCPs.

In l

addition to the conditions of question 4.17b, an operating level of 50% would be used if MFW is available and automatically controlling

level, l

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Reference:

OP-OC-SPS-SY-EF, page 32 OP-OC-SPS-3Y-FDW, page 15 4.18 a An SRO licensed supervisor at Oconee is a redundant statement when referring to operations personnel.

By definition, at Octnee all SRO licensed operators are considered to be supervisors.

A statement that at least one of the operators must hold an SRO license should be sufficient to meet the supervisor requirement for Operation procedures.

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i (nrc1287.com 7) i

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a ATTACHMENT 1 4.18 b The sequence change is actually documented by the SRO when approving the change.

The procedure user's documentation will be accomplished by the use of sign-off steps with date and time (this would have been done regardless of sequence change).

Just the action of the SRO needs to be addressed in this answer.

Reference:

Both answers 4.18a and 4.18b are verbatim answers from the references provided on the key.

The clarification provided here involves an interpretation of these requirements and defines how these requirements are met in actual application at Oconee.

4.19 Additional answers to this question would be - 30% power, ~ 65%

power, and - 90% power during power escalation.

Reference:

Op/1-2-3/A/1102/04, Encl. 3.3 i

(nrc1287.com 8)

ADDITIONAL REFRENCE QUESTION 2.05 NSM-2113 OTSG Chemical Cleaning Magnetite deposits (FE 0.) have been circumstantially related to increasing pressure drop across the A and B steam generators The OTSG Chemical Cleaning System designed by NUS Corporation will be used to remove these deposits ut,ing ths solvent recommended by the EPRI Steam Generator Owners group and developed / qualified by B&W. The Chemical Cleaning System is designed to mix and heat the cleaning solvent as well as the solutions for rinsing and passivating the steam gene.tator. The system also injects and circulated the solutions and transfers them to temporary waste storage tanks.

NSM-2162 Replace existing Lonegran Relief valves 1HP-302 and 1HP-71.

This NSM replaces existing Lonegran Relief Valves (1HP-302 and 1HP-71) with new bellow type relief valves. Reroute inlet and outlet pip!ng and install new expansion joints on outlet side of each valve.

NSM-2178 Replace all Bailey Meter Co. "VR" Recorders in the control ras.

with newer design.

The purpose of this modification is for the replacement of the Bailey-Meter Co. "WR" Recorders in the Unit 1 Control Room, due to the Bailey Recorders being obsolete and unavailability of replacement of "WR" recorders or parts.

NSM-2320 Replace IRC-7.

The purpose of this modification is to replace piston operated containment isolation valve IRC-7 with a smaller valve and operator qualified to isolate against full system differential pressure.

ADDITIONAL REFRENCE QUESTION 3.19_

OP-OC-IC-RCI DECEMBER 11, 1984 JMB/

<((gg/kg 3 &7-//[

PAGE 37A OF 49 P4I onH7 fM

2. Calculations

a. Average Core Temperature from Incore Thermocouples

1) When Power is > 2%, the program takes the average of the operable Incores not being used by the SSF (47).

a) If any of the forty-seven are > 50 F higher or lower than the average it will not be used in the calculation.

b) If an Incore is in Scan Lockout or contains an inserted value it will not be used in the calculation.

2) When power is < 2% for ~ 45 seconds, the program takes the average of the five highest operable qualified Incores.

a) If any of the twenty-four environmentally qualified Incores are > 50 F higher or lower than the average of the twenty-four it will not be used in the calculation.

t c'

b) If any of the twenty-four are in Scan Lockout or contain an inserted value it will not be used in the calculation.

3) Use of only the qua..ified Incores is not mandatory unless 1

a hostile environment exists in the R.B.

a) If power is > 2% it is assumed that a hostile environment does not exist.

4) Use of the forty-seven Incores gives a more accurate subcooling Margin indication and will not lead to tripping the RCP's unnecessarily following a "normal" l

reactor trip.

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7 ATTACHMENT 2 COMMENTS REGARDING THE SENIOR REACTOR OPERATOR LICENSING EXAMINATION Category 5.0 Theory of Nuclear Power Plant Operation, Fluids, and Thermodynamics 5.15 The answer given for this question is incorrect.

There is no correct choice.

At 32%' reactor power the primary system contributes 80% of the Oconee heat balance.

The weighting factor is equal to.2 and the primary contribution is equal to (1.2) or 80%.

Reference:

Duke Power Thermodynamics of Nuclear Power Plants, Chapter 3, page 188 5.17 An additional correct answer would be the lower temperature water from the UST or sprayed directly on OTSG tubes.

Depending on how "not preheated" is interpreted.

Reference:

RO question 1.20 answer and comment 5.24 Another acceptable answer would address the change in the coefficient as fuel temperature changes.

Over core life the centerline fuel temp decreases due to pellet swell and clad creep causing the gap between the pellet and clad to decrease thereby increasing the heal transfer coefficient.

This has the effect of increasing the.nagnitude (FTC becomes more negative) of the Fuel Temperature Cojfficient.

Reference:

n?-OC-SPS-RT-RC, pages 10-12 Fundamentals of Nuclear Reactor Engineering, page 149 Category 6.0 Plant Systems Design, Control and Instrumentation 6.05 b&d Refer to R.O.

Licensing Exam 2.05 b&d 6.07 This question did not specify whether the ACTUAL actuation setpoint or the TECHNICAL SPECIFICATION actuation 7etpoint was desired.

Since no specification was made some candidates gave the T.S.

actuation setpoints.

The answer key cr.ly references the actual setpoints.

Other acceptable answers would include the T.S.

(nrc1287.com 9)

ATTACHMENT 2 actuation setpoints of:

a.

2 1500 psig RCS, 5 4 psig RB b.

2 500 psig RCS, 5 4 psig RB c.

5 4 psig RB d.

5 30 psig RB

Reference:

ONS TECHNICAL SPECIFICATION 3.5.3 6.12 This question caused confusion for the candidates.

Clarification provided still left confusion on what was expected for an answer.

The question seems to ask for the design features and purposes of the FGC system rather than the ductwork of the system.

Other correct answers to this question could include the heat removal capability of the ROCUs, RBCU fan starts / speed shifts on ES actuation, and LPSW flow through the RBCUs during ES.

Reference:

OP-OC-SPS-SY-RBC, pages 10 and 19 6.13 Refer to R.O. Licensing Exam 3.13 (MS-126 and MS-129 have incorrect name/ description) 6.18 Refer to R.O. Licensing Exam 3.24 6.20 a Rotating the switch from the "OPERATE" position will cause the associated analog channel to trip because the switch must pass through the "TEST OPEBATE" position when rotated.

6.20 b The "Output State" toggle switch may also be referred to as the l

"Reset" toggle switch.

6.21 a The answer key specifies only the "A"

MDEFWP.

The answer should be auto start of "any" MDEFWP ("A" and/or "B" MDEFWP) since the question is worded "FDW-315/316".

Reference:

OP-OC-SPS-SY-EF, page 26 (D.2.a & b) l l

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(nrc1287.com 10)

ATTACHMENT 2 Category 7.0 Procedures - Normal, Abnormal, Emergency and Radiological Control 7.08 An additional answer could involve the test or maintenance allowance for one of the two required RBC trains when the reactor core is-subcritical with RCS pressure 2 350 psig or RCS temperature 2 250*F.

Reference:

ONS Technical Specification 3.3.5 b(2) 7.16 The answer key only list interlocks.

Since the questions asked for "requirements", additional correct answers could involve administrative limits.

Prior to disengagement the hoist tape readings must be within an acceptable band and a stable count rate must be verified.

Also the interlocks on the Unit 3 Multifunction Mast are different from those on the answer key.

1.

No dillion requirement of < 1200 pounds 2.

Low load light on (800 pounds) 3.

Air pressure available (pneumatic system) 4.

Uses hooked fingers (no licking pins)

Reference:

OP/1-2-3/A/1502/07, Limits and Precautions 2.19 and 2.21 OP-OC-SPS-FH-FHB, pages 34 and 38 (part 17) drwg OC-FH-FHB-02B i

i 7.19 Refer to R.O.

Licensing Exam 4.17 7.20 a The Reactor Trip Recovery procedure provided with the examination did not have the step referenced on the answer key.

Apparently an outdated copy of the procedure was used for the attachment.

The answer key references a more recent, updated, procedure.

The "Where" portion of this question should be deleted.

Reference:

OP/2/A/1102/02, dated 5-16-86 (attached to exam)

OP/2/A/1102/02, dated 4-13-87 (referenced on key) l l

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(nrcl287.com 11) l l

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ATTACHMENT 2 Category 8.0 Administrative Procedures, Conditions, and Limitations 8.07 Choices a and c are correct.

The work group is allowed to place supplemental tags on station equipment that has a two part serialized red tag attached.

They may not place a supplemental tag if a two part serialized tag is not attached.

Reference:

Station Directive 3.1.1, page 3 (4.6.1 & 4.6.3) 8.11 This question did not specify Duke Admin. limits or 10CFR20 limits.

The answer key references only the Duke Admin. limits.

Additional acceptable answers would be a) 3000 MREM and b) 12000 MREM (QTR limit x 4 QTRs) if the 10CFR20 limits are used and it is assumed that 5(N-18) limit will'not be exceeded.

Reference:

10CFR20 (in Health Physics Manual)

DPC Health Physics Manual, page 10 8.13 b The answer key references only the county the school is in.

The name of the school is Keowee School (in Oconee County) and Daniel High School (in Pickens County).

Reference:

0113 Implementing Procedure, RP/0/B/1000/10, Encl. 42 1

8.15 asb The answer key is incorrect.

The time and rate given have been deleted from the OMP.

The current OMP requires only that the LCO be evaluated and that the unit meet the required time frames / conditions addressed in the LCO.

The revised OMP 1-4 was sent in the update package dated 10-5-87.

Reference:

ONS Operations Manual OMP 1-4, page 2 (attached) 8.16 asb Refer to R. O.

Licensing Exam 4.18 a&b 8.18 and 8.19 reference the ONS Emergency Plan.

The Emergency Plan is a theory based / big picture approach to handling an emergency situation.

In order to implement the Emergency Plan, Implementing Procedures were developed.

These procedures provide the specific actions required to carry out the Emergency Plan.

The Implementing (nre1287.com 12)

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ATTACHMENT 2 Procedures that affect operations can be found in the RP/0/B/1000 series.

A more appropriate answer / reference can be found in the Implementing Procedures.

8.18 b This question was confusing for the Sh0 candidates.

Any time the criteria for escalation or deescalation is met, the emergency status will be revised.

Dur4.ng the time frame it requires to establish the Crisis Management Center, the shift supervisor (SRO) will be relieved by the station manager as Emergency Coordinator.

The SRO, acting as the Emergency Coordinator, will not deal with the Crisis Management Center.

Reference:

ONS Implementing Procedures, Vol. C RP/0/B/1000/3, page 2 and Encl. 4.1 RP/0/B/1000/4, page 3 and Encl. 4.1 RP/0/B/1000/5, page 3 and Encl. 4.1 8.19 Additional answers should include state / county and NRC notifications.

t The Emergency Response organization on-site is established by initiating a Site assembly and making the necessary PA announcements.

However, off-site agencies must also be alerted to complete the Emergency Response Organization.

State / county agenciet must be notified within 15 minutes of classification and the NRC must notified within one hour of classification.

Reference:

ONS Implementing Procedures, Vol. C RP/0/B/1000/02, page 1 RP/0/B/1000/03, page 2 t

RP/0/B/1000/04, page 1&2 l

RP/0/B/1000/05, pages 1&2 8.21 b This question can be interpreted to be asking for an individual's responsibility with respect to dose control /ALARA.

With this interpretation additional acceptable answers would include minimizing dose by minimizing time of exposure, decreasing dose rate by use of shielding or low dose zones, referring to applicable l

RWP/SRWP, notifying HP prior to entry for HP support and other methods to achieve ALARA.

Reference:

DPC Health Physics Manual, pages 18, 19 & 20 l

(nrc1287.com 13)

ATTACHMENT 2 8.25 The bases given on the answer key is incorrect.

Technical I-specification 3.6.3.b.2 does not allow operation of the Reactor Building Purge.

This portion of T.S.3.6 allows one of two isolation. valves on each penetration to be opened for test or maintenance when the reactor is at or below hot shutdown with RCS temperature > 250 F and RCS pressure > 350 psig.

The redundant isolation valve must be closed.

The correct answer should address the operating margin provided above the NPSH for RCP operation while operating the purge.

Reference:

ONS Technical Specification 3.6, page 3.6-3 (nrcl287.com 14) 1

ADDITIONAL REFRENCE QUESTION 8.15 Reviewed by O/4l- [0M Approved by Date f//V//7 Revision #

1 OCONEE NUCLEAR STATION OPERATIONS MANAGEMENT PROCEDURE 1-4 ACTIONS TO BE TAKEN IN THE CASE OF EXCEEDING LIMITS 1.0 Purpose The purpose of this OMP is to:

1)

Describe the actions to be taken whenever limits are exceeded.

2.0 References Technical Specifications 3.0 Responsibilities 3.1 If any of the following incidents Occur, the reactor shall be s'r.utdown immediately and maintained in a safe shutdown condition until otherwise authorized by the Nuclear Regulatory Commission. The Superintendent of Operations and the Station Manager shall be notified immediately.

A.

A safety limit, as defined in Section 2 of Technical Specifications, is exceeded.

B.

Unplanned post-trip retriticality occurs.

C.

Post trip pressure response results in E.S. actuation or RCS PORV/ code safety valve opening.

D.

An RCS temperature transient results in exceeding PTS limits or a loss of subcooled margin.

E.

Pressurizer level goes off-scale low with a loss of subcooling margin or; pressurizer level goes off-scale high resulting in RCS PORV/ code safety valve opening.

F.

OTSG pressure exceeds 1155 psig (ASME Code limit).

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OMP 1-4 Pags 2 of 2 3.2 Where a Limiting Safety System Setting, as defined in Section 2.3 of the Technical Specifications, is reached or exceeded, the reactor will be placed in a safe shutdown condition and maintained shutdown until otherwise authorized by the Station Manager. The Superintendent of Operations and the Station Manager shall be notified immediately.

3.3 Where a Limiting Condition for Operation (as defined in Section 3 of Technical Specifications) requires a unit shutdown and/or cooldown, the rate of shutdown and/or cooldown is to be determined by Operations such that the required condition is achieved in a controlled manner within the time specified.

Under the LCO action statements in Section 3 of Technical Specifications, it is permissible to maintain the plant in a steady-state mode (power level, RCS press / temp, etc.) in anticipation of resolving the item (s) that placed the unit under the LCO.

In these cases, the plant conditions specified by the LCO action statement must be achieved in a controlled manner within the maximum time allowable under the restrictions of the LCO.

EXAMPLE:

If the action statement of an LCO states that the reactor must be at hot shutdown in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, but it is anticipated that repairs will be completed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> which will satisfy the LCO, then the unit may remain at its present power level in anticipation of the successful repairs. The unit should maintain steady-state conditions during the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> hold period.

If at anytime it is discovered that the repairs cannot be accomplished well enough within the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> time frame to ensure a controlled shutdown, the unit shall immediately be shutdown in order to satisfy the time limit of the action statement.

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ATTACHMENT 3 The Reactor Operator and Senior Reactor Operator Written Licensing Examinations were reviewed on the basis of applicability to a licensed operator's ability to perform his duties.

The review considered only the intent of each guestion and did not address the technical accuracy of the questions (technical accuracy has been previously addressed).

The results of this review are as follows.

Strongly Oriented to Operator 1.03/5.11 2.06 4.11/7.09 7.02 1.04 2.07 4.12/7.15 7.12 1.05 2.16 4.22/7.22 8.04 1.10/5.12 2.23 4.23/7.11 8.05 1.11 3.09/6.03 7.01 8.15 Operator Oriented 1.02 2.04 3.07 4.14 6.14 8.09 1.06 2.05/6.05 3.12(1) 4.15 6.15 8.10 1.07/5.03 2.09 3.13/6.13 4.17/7.19 6.20 8.11 1.08/5.05 2.10 3.16 4.21 6.21 8.13 1.09/5.06 2.11 3.18 5.04 7.07 8.14 1.18 2.12 3.22 5.13 7.08(1) 8.19 1.21 2.14 3.25/6.19 5.19 8.01 8.20 1.22 2.19(1) 4.01 5.20 8.02 8.22(2) 1.24/5.25 2.20 4.02 5.21 8.07 8.23 1.13/5.07 2.21 4.03/7.03 6.02 8.08 1.20/5.17 3.06 4.06 6.06 4.13 6.07 6.08 (1)

If reworded (2)

If given T.S.

Applies but does not test ability i

1.12/5.08 2.13 3.21 4.20 7.14 8.24 l

1.14/5.09 2.17 3.24/6.18 5.01 7.16 8.25 l

1.15/5.10 3.01 4.04 5.14 7.18 1.16/5.18 3.03 4.05 5.15 7.20 1.17 3.04 4.08/8.12 5.16 7.21 1.19/5.02 3.08 4.09/7.10 5.23 8.03 1.23/5.22 3.10 4.10 5.24 8.06 2.01/6.01 3.14 4.16/7.17 7.04 8.12 3.15 4.18/8.16 7.05 8.17 3.17/6.11 4.19 7.06 i

l 3.20/6.09 7.13 3.05/6.04 (nrc1287.com 15)

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ATTACHMENT 3 Not Operator Oriented 1.01 3.02 6.12 2.02 3.11/6.10 8.18(1) 2.03 3.19 2.08 3.23/6.17 2.18 4.07(1) 2.22 (1) b part only Beyond scope of OMP-2-1 2.01 3.23/6.17 8.22 4.04 7.13 8.24 5.01 7.14 8.25 7.05 7.18 7.06 8.03 6.12 8.06

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l l

l 1

l (nre1287.com 16)

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