ML20237J233

Forwards Compilation of Training Staff Comments on Written License Exams Administered on 870818,per NUREG-1021

ML20237J233
Person / Time
Site:	Millstone
Issue date:	08/20/1987
From:	Scace S NORTHEAST UTILITIES
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
RTR-NUREG-1021 MP-10745, NUDOCS 8708260075
Download: ML20237J233 (20)

Text

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Eb General offices

• Selden Street Berlin. Connecticut I

wrma wsucwers tacmc a""

P.o. BOX 270 ErIIu',.^ IuYc~E,i.,

HARTFORD. CONN ECTICUT 06141-0270 g

www w:a= v.o.ovcwe (203) 665-5000 MP-10745 Re: NUREG-1021/ES-201/ para H.1 U.S.

Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555

Reference:

Facility Operating License No. NPF-49 Docket No. 50-423 August 18, 1987 NRC License Examination Comments Gentlemen:

Attached is the compilation of comments on the written examinations administered to Millstone Unit No. 3 license candidates on August 18, 1987.

These comments were the result of a review of the examinations conducted by members of the Millstone Unit No. 3 training staff.

Included are both the comments discussed during the exam review meeting of August 18, 1987 plus additional comments resulting from reviews conducted subsequent to this meeting. Attendees at the August 18, 1987 meeting were:

D.

Silk - NRC R.

Martin - Northeast Utilities R.

Temps - NRC M.

Mochimann - Northeast Utilities The exam reviews were conducted considering the following:

1.

Does the question elicit the correct response?

2.

Is the key answer correct?

3.

Is there potential for additional correct responses?

4.

Is the question appropriate?

References are provided, where necessary, to substantiate the comments.

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8708260075 070820 PDR ADOCK 05000423 1

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I Please contact Mr. Ron Stotts, Supervisor, Operator Training, Millstone Unit No.

3, with any questions concerning our comments.

l Yours truly, j

NORTHEAST NUCLEAR ENERGY COMPANY i

M St phen E.

Scace Station Superintendent

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Millstone Nuclear Power Station j

SES/MJM:jas

Attachment:

Senior Retctor Operator Exam Comments e.7d applicable references cc:

S.

Collins, Branch 'hief, Region I B. W.

Ruth, Manager iperator Training i

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SENIOR REACTOR OPERATOR EXAM 5.

THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS AND THERMODYNAMICS 5.06a Agreed that the reason for shutdown margin increasing may also be that the control rods move further out resulting in more negative reactivity inserted on a reactor trip (Reference OP3209B-1) 5.07c Agreed that a discussion of head loss is not required for full credit on this question.

(Reference Reactor Theory HTFF 2 of 2, Boiling Process, Pg 25) 5.10.b Agreed that the answer key addresses the long term response and consideration will be given, on a case by case basis, for an explanation which includes Psteam initially increasing due to Thot increasing on a loss of feedwater to the steam generator.

5.10.c Agreed to change part 2 to "NO CHANGE" and accept both "NO CHANGE" and " INCREASE" as acceptable answers for part 4 due to the guidance provided to the operators in step 10 of EOP 35 ES-0.2.

6.

PLANT SYSTEMS DESIGN, CONTROL AND INSTRUMENTATION 6.08 Agreed that answer s;tould be rods move in due to NI/ Pimp mismatch resulting in reactor trip on low PZR level (Reference-Instrument Failure Analysis Text pp 45-46) 7 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 7.03a Agreed to accept any two permanent plant people to accompany the NRC inspector into containment.

(Reference OP3212-1) 1 of 2 L

I' 7.10c Agreed that an acceptable alternate answer may be as long as the SI termination criteria in EOP 35 E-3 are met the core is subcooled, regardless of the conditions in the reactor vessel head, and SI may be terminated.

(Reference EOP 35 E-3 step 20) 8.

ADMINISTRATIVE PROCEDURES, CONDITIONS AND LIMITATIONS 8.01 Agreed that acceptatle answer may include comment that concurrence of two licensed Senior Reactor Operators (SRO's) >ne of whom is the onshift Shift Supervisor is nece ssary to implement a non-intent change.

(Reference - ACP 3.02 p.

30) 8.05 Agreed that acceptable answer for (a) may be shift crew complement may be one less than minimum required for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided immediate action taken to restore shift crew to within minimum requirements and that if this answer given for (a), acceptable answer for (b) may be in accordance with Tech Spec. 3.03.

Also agreed that reference to ACP 1.19 in answer key is not applicable to this situation.

(Reference -

Technical Specifications, Table 6.2-1) 8.08b Agreed to accept Security Shift Supervisor as alternative answer (Reference-ACP 7.04 p.3) 8.09 Agreed that acceptable answer is whenever following guidance in emergency operating procedures, as question inquires only "under what circumstances" is it permissible to violate LCO, and does not request any further explanation.

(Reference-EOP Format and Use Text pp 12-13) i

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3-e m o FOSAPPROVEDBYUNIT3' SUPERINTENDENT EFFECTIVE DATE PORC MTG NO.

SHUTDOWN MARGIN FOR MODES 1, 2 Calculated By Approved By

Date, Time 1.

Current Conditions Power level from NI-41, 42, 43, or 44 Tavg 412

• F 422

• F Tref TR412 432

• F 442

• F Rod Positions (DRPI or Step Counters) (circle one)

• - CD' steps SDE steps CC steps SDD steps

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CB steps SDC steps CC steps SDB steps SDA steps Indicate any Stuck, Dropped, or Misaligned Rods 1

i RCS Borc, Concentration j

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OPS Form 32098-1

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Rev. 0 Page 1 of 2 L4... "

CAUTION: Write in positive numbers only.

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2.

Total Rod Worth pcm l

3.

Add the following:

'pcm 3.1 Integrated rod worth X

-1

=

(of any groups. inserted into core) 3.2 Dropped / Misaligned rods s

X 200 pcm =

pcm Nu 'ier of Rods

,,g.. $ ;,3.3 Stuck Rods a

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X 1400 pcm =

pcm Number of Rods 3.4 Power Defect X -1

=

pcm 3.5 Moderator Correction X

=

pcm Tref-Tavg MTC 3.6 Total pcm 4.

Step 2 - Step 3.6 pcm i

5.

Subtract 4000 pcm I

6.

Total pcm

-NOTE:

A positive value indicates adequate SHUTDOWN MARGIN.

i Forward to Reactor Engineering when completed.

Reviewed By:

Date:

f Reac, tor Engineering OPS Form 3209B-1 Rev. 0 Page 2 of 2

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BOILING PROCESSES g

Instructor Lesson Content Instructor /

Aids Student Activity.

5.

Axial Shape: If power is peaked to the top of the core, this puts that region closer to DNB.

6.

Radial Shape - distortions in the radial shape can cause some fuel assemblies I

to be hotter than average, f

which is closer to DNB.

d.

Departure from Nucleate Boiling Ratio (DNBR) 1.

DNBR is a convenient way to express the margin to DNBR.

l 2.

Defined: DNBR = critical heat flux / actual heat flux 3.

DNBR: safer is higher, bigger is better 4.

Licensing Value: 1.3 or more i

Page 25 of 31 L

NATURAL CIRCULATION COOLDOWN E0P 35 ES-0.2 Page 5 Rev. I k

STEP

. ACTIONMIPECTED MESPONSE RESPONSE NOT OBTAINED NOTE To reduce thermal shock on the RCS if there is no urgency to immediately depressurize the RCS, PZR heaters may be de-energized to allow the RCS to depressurize slowly by heat losses to ambient.

8 Depressurize RCS To l

approximately 1950 PSIA a.

Check letdown - IN SERVICE a.

Try to establish letdown.

IF letdown can NOT be established, THEN use L

one PZR PORV.

Go to Step 9.

b.

Use auxiliary spray CAUTION SI actuation circuits will automatically unblock if PZR pressure increases to > 2000 PSIA.

f 9

Block SI Actuation train A and B a.

Main Steam Line SI -

Block A/B b.

PZR SI - BLOCK A/B 10 Maintain Following RCS Conditions a.

RCS pressure - AT approximately 1950 PSIA b.

PZR level - AT approximately 25%

c.

Cooldown rate in RCS cold legs - < 50'F/HR d.

Maintain RCS temperature and pressure to the right of the 60'F/HR curve on Tech Spec Figure 3.4-3 RCS*C00LDOWN LIM 1IATIONS I

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7. Power range indications (NI 41C, 42C, 43C, 44C); MB4; meter readings increasing

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8.; Tavg/auc ref recorder (TR 412); MB4; if initially at lower

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' power levels, Tref _ pen failed to full-load.value; T,yg pen increasing in response to rod motion 9.

PT 505 meter indication (PI 505); MB7; failed high The'generaloperatoractions.tobejakenonahighfailureofimpulsepressure channel 505 are the.following:

1.

Upon noting rods moving.out at high speed, place rod control in manual. zutilizing manual rod control, lower reactor power below secondary plant load to restore T,yg.to the appropriate. reference value.

2.

Select PT 506 as the input to the automatic.SGWLC S/ stem and Rod Control System.

3.

Place rod control in automatic.

4, The steam dump load rejection controller is no longer available due to the failure of PT 505. Therefore, the steam dumps must be placed in the STEAM PRESSURE mode.

Check PK 507 A/M station on MSS.

Ensure the automatic setpoint is adjusted to the no-load value of 1092 psig.-

Select the steam dump MODE SELECTOR switch to'the STEAM PRESSURE mode.

The arming solenoid valve is now armed. The dump valves'will respond to differences between actual and no-load steam pressure.

A low failure of channel 505 forces the plant to respond in a manner that is opposite to that presented in the high-failure case. The T circuit fails ref

-to its low limit of 557 4 A large error is seen by the rod control temperature mismatch circuit. The P low failure is seen by the rate of 4gp change of power mismatch circuit as turbine power dropping below reactor power 45 06195:4

at a high rate of. speed. Both signals combine to cause rods to drive inward

~ at the maximum rate of speed.

Inward rod speed will die off as the following occur:

1. -The rate characteristics of the power mismatch circuit cause the circuit output to drop.

due to rod 2.

Auctioneered T,yg begins decreasing toward Tref insertion (no effect until T,yg is within 5'F of Tref)'

3.

Auctioneered reactor power drops due to rod insertion.

Note that automatic rod withdrawal is prohibited by C-5.

4.

As T,yg drops because of the mismatch between reactor power and steam Pressuri:er level initially drops rapidly, demand, the coolant will contract.

Since the low pressurizer pressure reactor trip as does plant pressure.

signal is " lead-lag" compensated, it is likely that the reactor will trip.

In addition, The reactor trip (p-4) will send a signal to trip the turbine.

because two of four T,yg channels are below the low T,yg setpoint of 564'F, a feedwater isolation will occur. This will essentially eliminate the mismatch between reactor output and steam plant output.

If reactor power had been at a lower initial level, such as 30 percent, a trip would most likely not have occurred. This is on low pressuri:er pressure at the initiation of the would have been within 8'F of Tref because T,yg Rod speed would begin tapering off as soon as auctioneered Tgg failure.

cropped by 3'F; In addition, the rate of change of power mismatch circuit would see a 30 percent step reduction vice a 100-percent step reduction.

If Higher rod speeds would thus be maintained for shorter periods of time.

reactor power should fall below 10 percent on 3 of 4 power range channels, a reactor trip would result. This is because the P-10 circuit would auto-matically reinitiate the intermediate range high flux and power range low Steam demand wculd then drag reactor power back setpoint reactor protection.

up to 25 percent, tripping the reactor.

j 46 06195:4

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_3-8 7-3 e Form Approved By Unit 3 Superintendent Effective Date PORC Mtg. No.

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CONTAINMENT ENTRY CHECK-OFF LIST 1.

Job Description

-1.1 Reason for entry 1.2 AWO #

1.3 RWP #

1.4 Estima'ted Duration

1. 5 List personnel to enter containment and their assignments below.

Each entry team must have their own observer, team leader, and hatch operator.

The same person may fill more than one position.

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NAME (Please Print)

Team Leader Hatch Operator Observer Other Other Other Other Other Department Supervisor Date

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OPS Form 3212-1 Rev. O Page 1 of 4

f' 2.

Medical Department Permission for Entry for Personnel Listed in Section 1 Initials

^2.1 Personnel listed for entry are qualified for E

containment entry as specified in Figure 7.3.

2.2 Personnel listed for entry have satisfactorily completed the Unit 3 Containment Entry Medical Checklist (SF 630).

Permission to enter containment is granted.

Medical Department Rep. Signature Date 3.

Team Leader Final Pre-Entry Check-Off Initials 3.1 All required tools and equipment necessary to perform task have been assembled.

3.2 Spare biopak is available to be carried into containment with the entry team.

3.3 Conditions expected in Containment including air sample results, (SF 3212-2) Wet Bulb Globe l

Temperature (WBGT) expected at the work site and l

proposed stay times have been discussed with the entry team 3.4 OP 3212, Figure 7.3, Subatmospheric Containment Entry Briefing, has teen reviewed with the entry team.

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OPS Form 3212-1 l

Rev. O Page 2 of 4

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Initials 3.5 The Observer is familiar with establishing stay

' times for hot areas as provided for in Figure 7.3 and is prepared to establish stay times for the workers.

3.6 The entry team has a knowledgeable hatch operator who is not a worker or at least two workers are knowledgeable hatch operators.

The operators are identified to the rest of the a

entry team.

3.7 Hand signals provided in OP 3212, Figure 7.4-have been reviewed with the entry tean,.

Team members understand the task, their assignments and the precautions related to Heat Stress and a Subatmospheric Contrain:aent.

Team Member Signsture Date l

I Team is ready for containment entry.

Team Leader Signature Date

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OPS Form 3212-1 Rev. O Page 3 of 4

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STEAM GENER4 TOR TUBE RUPTURE E0P 35 E-3 Pago 15

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Rev. 1 STEP-

- ACTION / EXPECTED RESFMSE RESPONSE NOT OSTAlWED.

CAUTION Voiding in the upper head region should NOT preclude SI termination. SI NUST BE TERMINATED when termination criteria are satisfied to prevent overfilling of the ruptured SG(s) l 20 Check If ECCS Flow Should Be Terminated o

a.

RCS subcooling based on a.

DO NOT STOP ECCS PUMPS.

core exit TCs - GREATER GO 70 E0P 35 ECA-3.1, THAN 30*F [90*F FOR SGTR WITH LOSS OF ADVERSE CONTAINMENT]

REACTOR COOLANT -

SUBC00 LED RECOVERY DESIRED, Step 1.

b.

Secondary heat sink b.

E neither condition satisfied, TEN DO NOT

1) Total feed flow to STOP ECCS PUMPS. Go to SG(s) - GREATER THAN EOP 35 ECA-3,1, SGTR

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525 GPM AVAILABLE WITH LOSS OF REACTOR COOLANT - SUBC00 LED OR RECOVERY DESIRED, Step 1.

2) Narrow range level in at least one intact SG -

GREATER THAN 4%

[34% FOR ADVERSE CONTAINMENT]

c.

RCS pressure - STABLE OR c.

DO NOT STOP ECCS PUMPS.

INCREASING Go to EOP 35 ECA-3.1, j

SGTR WITH LOSS OF REACTOR COOLANT - SUBC00 LED RECOVERY DESIRED, Step 1.

1 d.

PZR level - GREATER THAN d.

DO NOT STOP ECCS PUMPS.

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7% [50% FOR ADVERSE Return to Step 13.

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CONTAINMENT]

21 STOP ECCS Pueps And Plcce In l

AUTO l

a.

SI pumps

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All but one charging pump

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!ACP-QA-3.02 Page'30 RGv. 40 1

Chances to Procedures & Forms _

6.9 Non-Intent Changeg 6.9.1.

Preparation 6.9.1.'1

_A non-intent change may be prepared by any member of the plant staff using the Station

^

Procedure or Forms Change Form, SF 302.

Numbering 6.9.1.2 Non-intent changes are to be_ sequentially numbered against the effective revision of SF 313 may be used the procedure or form.

as an index by departments that require it.

Approval and Implementation

6. 9.1. 3 The concurrence of two licensed Senior Reactor Operators (SRO's) from the particular unit involved is necessary to approve and implement a non-intent change At least one of the on that unit.

individuals shall be the on duty Shift The concurrence of two Supervisor.

j licensed Senior Reactor Operators per unit is necessary to approve and implement a Common Site Station Services Procedure or form change, e.g., Security Procedure change. At least one of the individuals per unit shall be the on duty Shift Supervisor.

This shall be documented in Section D of SF 302.

Final Approval 6.9.1.4 All non-intent changes for SORC/PORC approved procedures and forms shall be forwarded through the responsible Department Head to SORC/PORC for review and to the Station / Unit Superintendent for L-approval, or to the Station / Unit

4 TABLE 6.2-1 I..

MINIMUM SHIFT CREW COMPOSITION l 1

. POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODE 1, 2, 3, or 4 MODE 5 or 6

~

SS 1

1 SRO 1

None R0 2

1 PE0 2

1 STA 1*

None Shift Supervisor with a Senior Operator license on Unit 3 SS Individual with a Senior Operator license on Unit 3 SRO Individual with an Operator license on Unit 3 RO Plant Equipment Operator (Non-licensed)

PE0 Shift Technical Advisor STA The shift crew composition may be one less than the minimum requirements of Table 6.2-1 for a period of time not to exceed.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.

This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

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During any absence of the Shift Supervisor from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with a valid Senior Operator license shall be designated to assume the control room command function.

During any absence of the Shift Supervisor from the control room while the unit is in MODE 5 or 6, an individual with a valid Senior Operator license or Operator license shall be designated to assume the control room command function.

I

• The STA position may be filled by an on-shift Senior Reactor Operator only if that Senior Reactor Operator' meets the Shift Technical Advisor qualifica-k tions of the Commission Policy Statement on Engineering Expertise on Shift.

MILLSTONE - UNIT 3 6-5 b'

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ACP 7.04 Page 3 Rev. 18 4.6 y.A.P. - Vehicle Access Point located in an east section of the protected fence.

5.

RESPONSIBILITY 5.1 _ Station Superintendent - Responsible for ensuring a program exists for the control of security locks and keys.

5.2 Station / Unit / Station Services Superintendent Responsible for authorizing rotation of security keys and locks as outlined in this procedure.

5. 3 _ Security Supervisors 5.3.1 Responsible for ensuring that security locks and keys are issued only when duly authorized.

5.3.2 Responsible for ensuring that security keys and locks are controlled when not issued.

5.4 _ Authorized Personnel 5.4.1 Responsible for using security keys only during periods when the security system is inoperative or g

under emergency conditions.

5.4.2 Responsible for reporting lost or broken keys and the locations of defective security locks to the Security Shift Supervisor.

5. 5. Security Shift Supervisor 5.5.1 Responsible for authorizing the issue of security keys during emergency conditions.

5.5.2 Responsible for authorizing the issue of security keys for surveillance testing / maintenance and when needed for Protected and Vital Area Access.

5. 6 Operating Shift Supervisor 5.6.1 Responsible for authorizing the issue of security keys during emergency conditions.

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the tree would indicate a SG in either wet layup or in dry layup to be abnormal.

To clarify the usability of the EOPs for transients originating during other than the assumed initial operating MODES, a detailed review of the entire network has been performed.

-A statement has been added at the beginning of the ENTRY CONDITIONS to inform the operator of the operational modes the procedure is applicable in.

If not in one of these modes, the operator will have to review each step to determine if it is still applicable for the present plant conditions.

2.5 EOPs IN RELATION TO TECHNICAL SPECIFICATIONS The plant Technical Specifications contain the limiting conditions for plant normal operation in the applicable modes.

By abiding by these conditions, the plant's operation would be conducted in a safe manner and the design safety features would be ready to respond if a design basis accident were to occur.

One could consider that the Technical Specifications play a preventative and preparative role in ensuring plant safety.

If an accident does occur, the Emergency Operating Procedures play a responsive role in dealing with the accidents.

The Emergency Operating Procedures provide the actions to be performed and parameters to be monitored to maintain plant safety and to achieve optimal recovery.

I When the safety systems are actuated and performing their j

role during an accident, any number of the preparative I

Technical Specifications will be violated due to the action of the safety systems (e.g., after the RWST is injected into bS_

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the RCS, its level will be below its Technical Specification

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limit).

The accident itself could be a violation of the preventative Technical Specifications (e.g., a LOCA will exceed RCS leak limits).

The Emergency Operating Procedures were developed to respond to accident conditions and are supported by extensive analytical background, in most cases best estimate.

The actions delineated in the EOPs are those actions necessary to deal with the accident in order to maintain or restore the plant in a safe condition.

In general, the Technical Specification limitations are considered in developing the emergency response actions in the EOPs.

However, the EOPs contain actions which will lead to Technical Specification violations in order to maintain plant safety (e.g., opening pressurizer PORVs during a complete loss of secondary heat sink will violate RCS leak limitations, but is necessary to

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provide for core cooling and prevent more severe consequences).

Although it is desirable to remain within Technical Specification limits at all times, one must keep in mind that the overall objective is to protect the health and safety of the public.

This may require violating a particular Technical Specification in response to an accident.

2.6 FOLDOUT PAGES The concept of the " Foldout Page" is utilized in the Emergency Operating Procedures to remind the operator that if I

certain parameters or conditions exceed their limitations, then the specified actions on the FOLDOUT PAGE should be taken.

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