Text

Licensee Post Exam Comments & NRC Resolution (Folder 1)
	ML051390229
Person / Time
Site:	Millstone
Issue date:	03/31/2005
From:	Price J; Dominion Nuclear Connecticut
To:	D'Antonio J; NRC/RGN-I/DRS/OSB
	Conte R
References
	04-570F
	Download: ML051390229 (136)
	v • d • e

Mr. Joseph M. DAntonio Serial No. 04-57OF Senior Examiner/lnspector 05 1% -7 7s 51 MPS Lic/BAK RO U.S. Nuclear Regulatory Commission Docket No. 50-336 Washington, DC 20555 License No. DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2 POST EXAMINATION ACTIVITIES FOR THE MARCH 2005 REACTOR OPERATOR AND SENIOR REACTOR OPERATOR INITIAL WRllTEN EXAMlNlATlONS In accordance with NUREG-1021, Examination Standard 501, Paragraph C, Section 1a, Dominion Nuclear Connecticut, Inc. hereby transmits supporting documentation, including post examination analysis, for Millstone Unit 2 Reactor (RO) and Senior Reactor Operator (SRO) Initial Written Examinations, administered on March 18, 2005.

Also included within this submittal is a request to accept two correct answers for Questions 32, 38, 66, 71 and 72 of the RO examination. Also, additional information is provided to support the acceptable performance of JPM A2SRO - SRO AWO Acceptance. transmits the examination analysis. Attachment 2 transmits technical justifications for accepting two answers for Questions 32, 38, 66, 71 and 72 of the RO examination and the additional information to support the acceptable performance of JPM A2SRO - SRO AWO Acceptance. As applicable, the technical justifications were reviewed by system experts and have management concurrence and approval. transmits the remaining required documentation.

If you have any questions or require additional information, please contact Mr. David W. Dodson at (860) 447-1791, extension 2346.

J. Alan Price Site Vice President - Millstone

Serial No. 04-570F Post Examination Activities Page 2 of 2 Attachments: 2

Enclosures:

1 Commitments made in this letter: None.

Cc: (w/o enclosure):

Mr. R. J. Conte Chief, Operational Safety Branch U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415 U.S.Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415 Mr. V. Nerses Senior Project Manager US. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8C2 Rockville, MD 20852-2738 Mr. S. M. Schneider NRC Senior Resident Inspector Millstone Power Station U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555

Serial No. 04-57OF Docket No. 50-336 Attachment 1 March 2005 Reactor Operator and Senior Reactor Operator Examination Analvsis Millstone Power Station 2 Dominion Nuclear Connecticut, Inc. (DNC)

Millstone Unit 2 LOlT 2005 RO and SRO Exam Analysis (Questions with at least a 50% miss rate)

Note: The numbers to the left of each letter (answeddistractor) indicate the number of candidates that selected that answer. The correct answer is circled.

3 of 7 ROs missed; 2 of 2 SROs missed

.# 16 Plant is operating in MODE 3. 'C' Service Water Pump is out of service for motor maintenance. 'B' Service Water Pump is supplying the 'B' Service Water header. Bus 24E is aligned to Bus 240. 'B' RBCCW Heat Exchanger is aligned to provide minimum flow for the ' A Service Water header. Long Island Sound water temperature is 37°F. EDG SW Bypass Valves SW-231A and SW-231 B are being maintained closed because of an issue with the adequacy of valve actuator spring pressure.

125 VDC Panel DVIO de-energizes due to a fault. You now have the following indications:

PPC Points:

F6433 'A' RBCCW HX SW Flow => 8885 gpm F6434 'B' RBCCW HX SW Flow => 795 gpm F6435 'C' RBCCW HX SW Flow => 8965 gpm Local EDG SW Flows:

FIC-6397 'A' EDG => 850 gpm FIC-6389 'B' EDG => 100 gpm Assuming NO operator actions related to Service Water, which of the following is correct concerning these indications?

3 A 'B' EDG SW flow is LOWER than normal because SW-89B, DG TCV is closed B 'A' EDG SW flow is HIGHER than normal because of a rupture downstream of 'A' EDG SW flow transmitter.

' c 'A' RBCCW HX SW flow is LOWER than normal because of diversion of flow to the in-service TBCCW HX.

'C' RBCCW HX SW flow is HIGHER than normal because of a rupture downstream of 'C' RBCCW HX SW flow transmitter.

The candidates that chose "A" knew that the 'B' EDG Service Water flow was lower than normal, but failed to recognize the correct reason for the lower flow (Rupture in "B" Service Water header). The remaining students were confused by the indications resulting from the loss of DV-10. The question i s technically accurate, but somewhat confusing due to two simultaneous malfunctions. Only one of the malfunctions is addressed in the question stem; the other must be obtained from the correct

iiisi\: er. This questions was covered during the post exam review. No modification is required.

Page 2 of 17

Millstone Unit 2 LOlT 2005 RO and SRO Exam Analysis (Questions with at least a 50% miss rate) 5 of 7 ROs missed

26 Unit 2 is operating at 100% power. Which of the following Unit 2 activities/events requires direct notification of Unit 3 personnel?

2 A planned release of Waste Gas Decay Tank T-19A 3 B entry into 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS action statement for a RWST boron sample result of 1685 ppm c manageable steam leak on body of HD-I03A, Feedwater Heater 1A Normal Dump Valve small oily rag bin fire in turbine building, extinguished within 10 minutes The candictates that chose B assumed entry into a one hour action statement would result in a plant shutdown. A reduction in power requires notificdtion of Unit 3. The candidates that chose A mistakenly thought that a waste gas discharge required notification of Unit 3. A call to Unit 3 to report the initiation of a waste gas discharge is sometimes performed as a courtesy. This question is technically correct, but distractor B is somewhat misleading if the candidate assumes the plant will be shut down due to entry into a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action statement. This questions was covered during the post exam review. No modification is required.

Page 3 of 17

Millstone Unit 2 LOlT 2005 RO and SRO Exam Analysis (Questions with at least a 50% miss rate) 6 of 7 ROs missed

32 The plant is operating at full power with all equipment functional, except for the B HPSl Pump, which IS 00s for maintenance.

Then, a large break LOCA occurs combined with a loss of Bus 24D (due to an electrical fault on 24D)

Which one of the choices correctly completes the following statement regarding the impact of the loss of ECCS pumps.

10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months after the event, a loss of the only available adversely affect long term core cooling because the remaining A HPSl pump would, LPSl pump does NOT have a system flowpath for boron precipitation control.

HPSl pump would, LPSl pump could NOT be procedurally realigned for boron precipitation control via hot leg injection.

c LPSl pump would NOT,HPSl pump is preferred for boron precipitation control D LPSl pump would NOT, HPSl pump could be procedurally realigned for boron precipitation control via hot leg injection.

Up011 relriew, it was noted that both A and B are correct answers. (See attached justification.) It is recommended that both A and B be accepted as correct and that this question be modified per the attached for future use.

Page 4 cif 17

1. 32 1 d RO id SRO Origin New 1 Memory? (Check=Yes)

The plant is operating at full power with all equipment functional, except for the 'B' HPSl Pump, which is 0 0 s for maintenance Then, a large break LOCA occurs combined with a loss of Bus 24D (due to an electrical fault on 24D)

Which one of the choices correctly completes the following statement regarding the impact of the loss of ECCS pumps.

10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months after the event, a loss of the only available adversely affect long term core cooling because the remaining A LPSl pump would, HPSl pump must be used for boron precipitation control instead of long term core cooling B LPSl pump would, HPSl pump can NOT be used for both boron precipitation control and long term core cooling c LPSl pump would NOT, HPSl pump can be used for both boron precipitation control and long term core cooling D LPSl pump would NOT, HPSl pump could be procedurally realigned for boron precipitation control via hot leg injection Justification CHOICE (A) - NO WRONG: LPSl is needed for Boron Precip. control.

VALID DISTRACTOR: However, HPSl injection is necessary for adequate injection flow to prevent core damage.

CHOICE ( B ) -YES A single HPSl pump will provide sufficient flow for long term cooling, but it cannot also be physically aligned for hot leg injection.

CHOICE (C) - NO WRONG: Loss of LPSl would have adverse affect because of inability to realign HPSl VALID DISTRACTOR: HPSl does provide sufficient core cooling flow.

CHOICE (D) - NO WRONG. HPSl could not be procedurally realigned for boron precipitation control because it is needed for injection.

VALID DISTRACTOR: HPSl could physically realigned but not iaw procedure.

Reference [006ECC-01-C RO-51, NRC-2005 Question Reworded to streamline stem and clarify distractors based on procedure requirements and system capabilities.

History NRC K/A System/E/A NRC K/A Generic System 006 Emergency Core Cooling System (ECCS)

Number K6.13 Knowledge of the effect of a loss or malfunction on the following will have on the ECCS: Pumps I m portance ROlSRO 2.8 3.1 IOCFR Link (CFR: 41 7 I 4 5 7)

Page 36 of 101

Millstone Unit 2 LOlT 2005 RO and SRO Exam Analysis (Questions with at least a 50% miss rate) 6 of 7 ROs missed; 1 of 2 SROs missed

38 Given the following plant conditions:

- 100% power

- SG levels at setpoint

- Steam flow and feed flow matched

- SG2 Feed Flow Transmitter FT-5269,, output fails high With NO operator actions, which of the following describes the expected plant response?

SG level lowers, but stabilizes above the low level reactor trip.

2 0 5 B SG level lowers to the low level reactor trip.

1 c SG level rises, but stabilizes below the high level turbine trip.

I D SG level rises to the high level turbine trip Upon review. it was noted that both A and b niay be considered correct answers. (See attached j~istification.)It is recommended that both A and B be accepted as correct and that this question be modified per the attached for future use.

Page 5 of 17

1. 38 1 v RO 9iSRO Question ID: 5000034 -1 ' 1 Memory? (Check=Yes) ,

Given the following plant conditions:

- 100% power

- SG levels at setpoint

- Steam flow and feed flow matched

- SG2 Feed Flow Transmitter FT-5269A output fails high With NO operator actions, which of the following describes the expected plant response?

A SG level will lower when the Feed Reg. Valve fully closes on the flow mismatch.

B SG level will lower on the flow mismatch, but stabilize above the low level RPS trip.

C SG level will rise on the feed flow dominant signal until the high level turbine trip.

D SG level will rise on the feed flow demand, but stabilize below the high level turbine trip.

Justification CHOICE (A) - NO WRONG: Output from feed flow transmitters FT-5269A and FT-52698 on the SG2 feed line are averaged for input to the three-element level control. Failing one transmitter high drives the average high The control system will respond by closing the FRV. The level signal will slowly act on the steam flow signal to moderate the response. However, because of the relatively small error of the feed flow signal and the fact that the signal is STEAM flow dominant, will result in little effect on the level input signal.

VALID DISTRACTOR: Applicant may think that feed flow signal will dominate level control, resulting in SG level continuing to lower.

CHOICE (6)-YES CORRECT: Output from feed flow transmitters FT-5269A and FT-5269B on the SG2 feed line are averaged for input to the three-element level control. Failing one feed flow transmitter high drives the average up a small amount. The control system will respond by closing the FRV. The level signal will slowly act on the steam flow signal to moderate the response. Without operator action, level will decrease a small amount, but soon be turned by both level error and steam flow dominance.

CHOICE (C) - NO WRONG: SG will lower to the low level trip setpoint.

VALID DISTRACTOR: Applicant may think that the feed flow error dominates the level control system, causing SG level to rise to the turbine trip.

CHOICE (D)- NO WRONG: SG will lower to the low level trip setpoint.

VALID DISTRACTOR: Applicant may think the higher indicated feed flow will cause SG level to rise but stabilize below the high level trip based on input to the control system from the level and steam flow signal.

Reference (059 FWC-01-C R 0 9 8858),NRC-2005 NRC KIA SystemlEIA NRC WA Generic System 059 Main Feedwater (MFW) System Number K4.08 Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following:

Feedwater regulatory valve operation (on basis of steam flow, feed flow mismatch)

Importance ROlSRO 2.5 2.7 IOCFR Link (CFR 41.7)

Page 68 of 176

Millstone Unit 2 LOlT 2005 RO and SRO Exam Analysis (Questions with at least a 50% miss rate) 4 of 7 ROs missed; 1 of 2 SROs missed

48 The plant is operating at 100% power when a large break LOCA occurs.

If malfunctions prevent the use of either hydrogen recombiner, identify the approximate time that will elapse from the start of the event before hydrogen concentration will reach 3% by volume inside j a contain me nt .

greater than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 2 B 48 to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months j c 24 to 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months D less than 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months This question is technically accurate. Knowledgc deficiencies were covered during the post exam revieu.. No modification is required.

Page 6 of 17

Millstone Unit 2 LOlT 2005 RO and SRO Exam Analysis (Questions with at least a 50% miss rate) 5 of 7 ROs missed

62 The plant is operating at 100% power when you receive a high radiation alarm on RM-8132A (Unit 2 Stack Particulate) .

Given the following events and conditions:

-When checked on the PPC and RC-14, RM-8132A is found to be reading 7.5E04 cpm and stable

- An air sample by Health Physics confirms that this is a valid alarm

- HP recommends revising the RM-8132A module setpoint(s) on RC-14 Using the provided attachment, which one of the following describes what must be done with the RM-8132A module setpoints?

A Raise the alarm setpoint Do NOT change the alert or fail setpoints 4 B Raise the alarm setpoint and the alert setpoint Do NOT change the fail setpoint Raise the alarm, alert, and fail setpoints D Do NOT change the alarm setpoints This question is technically accurate. The candidates that chose "B" did not realize that the fail setpoint needed to be changed to alert the operator to changing (lowering) radiological conditions. If the radiation levels subsequently lower.(after the alert and alarm values are raised), then the alert and alarm setpoint will remain at the higher value and will be less conservative. Knowledge deficiencies were covered during the post exam review. No modification is required.

Page 5 of 17

Millstone Unit 2 LOlT 2005 RO and SRO Exam Analysis (Questions with at least a 50% miss rate) 7 of 7 ROs missed; 2 of 2 SROs missed

64 Unit 2 is in MODE 5 preparing to commence a refueling outage. Given the following events and conditions:

- A containment purge is being lined up in accordance with OP-23148, "Containment and Enclosure Building Purge" using the containment cleanup mode

- Noble gas concentration inside containment exceeds the limits in OP-2314B Which one of the following statements correctly describes the purge path required?

Containment purge flow will be directed through the EBFAS system through 2-AC-3 (EB PURGE SUPPLY DMPR)

Containment purge flow will be directed through the EBFAS system through 2-AC-57 (CTMT PURGE EXH DMPR)

C Containment purge flow will be directed through the main exhaust system through 2-AC-3 (EB PURGE SUPPLY DMPR)

D Containment purge flow will be directed through the main exhaust system through 2-AC-57 (CTMT PURGE EXH DMPR)

This question is technically accurate; however it requires the candidates to remember the specific flowpath for a ventilation alignment that is rarely used. Because this flowpath is rarely used. the nomenclature for the damper listed in "B" (CTMT PITRGE EXH DMPR) coupled with the evolution described in the stem (containment purge.. .using the containment cleanup mode) caused confusion.

Knowledge deficiencies were covered during the post exam review. No modification is required.

Page 8 of 1'7

Millstone Unit 2 LOlT 2005 RO and SRO Exam Analysis (Questions with at least a 50% miss rate) 7 of 7 ROs missed; 2 of 2 SROs missed

66 Unit 2 is conducting a reactor start up. Given the following events and conditions:

- Wide range (WR) logarithmic nuclear instrument (NI) channels C and D are out of service

- The reactor is not yet critical

- The ECP expected critical rod height is 100 steps on Regulating Group 6

- Regulating Group 4 is withdrawn to 60 steps

- WR NI Channel A failed low WRL NI Channel A <I .OE-I CPS WRL NI Channel B 6.2E2CPS Which one of the following statements correctly describes the required action (if any) required to comply with TECHNICAL SPECIFICATIONS?

A Immediately trip the reactor.

4 B Insert all control rods and shutdown the reactor 5 c Stop the startup until WRL NI Channel A has been repaired. NO other actions are required.

a Immediately ensure adequate shutdown margin IJpon revie\v, it was noted that both B and D are correct answers. (See attached justification.) It is recommended that both B and D be accepted as correct and that this question be modified per the attached for future use.

Page 9 of 17

1. 66 ~] v RO vi SRO Question ID: 500006g 1 7 Memory? (Check=Yes)

A reactor start up is in progress with the following conditions

-Wide range (WR) logarithmic nuclear instrument (NI) channels 'C' and 'D' are out of service

- The reactor is not yet critical

- Regulating Group 4 has just been withdrawn to 80 steps.

Then, Wide Range NI Channel ' A fails low, resulting in the following indications on the two Wide Range Channels being monitored:

- WRL NI Channel 'A' <I .OE-1 CPS

- WRL NI Channel 'B' 6.2E2 CPS

- Startup Rate for Channel 'B' is 0.1 DPM Which one of the following statements correctly describes the required action based on this instrument failure?

A SIMULTANEOUSLY insert all Shutdown and Regulating control rods immediately.

B Immediately MANUALLY insert all Shutdown and Regulating control rods.

C IMMEDIATELY stop the startup until Wide Range NI Ch. ' A has been repaired. v' D IMMEDIATELY commence Emergency Boration to Cold Shutdown concentration.

Justification CHOICE [A] - NO WRONG This action would only be required if startup rate exceeded the limits stated in OP-2202 CHOICE [B] - NO WRONG This action would only be required if the reactor was believed to be going critical far off from the CEA level predicted by the ECP. With the remaining WR NI reading only 620 cps at this rod height, this is unlikely.

CHOICE [C] - YES CORRECT - Although the plant is "officially" in Mode 2, and the requirements of Tech Spec 3.3.1.1 (2 WRL NI channels must be operable) do not apply per Tech. Specs., the Reactor Startup procedure (OP-2202)

DOES require two WR NI channels be operable to continue the reactor startup. Therefore, stopping the startup is the required action.

CHOICE [D] - NO WRONG The reactor has not yet transitioned to mode 2 and 2 WRL NI channels are required in mode 3.

With only 1 WRL NI channel operable, the tech spec action is to immediately determine shutdown margin.

Reference NRC-2005 NRC KIA System/E/A NRC K/A Generic System 033 Loss of Intermediate Range 2.2 Equipment Control Nuclear Instrumentation Number GA 2.2.22 SEE GENERIC WA Knowledge of limiting conditions for operations and safety limits Importance ROlSRO 3.4 4.1 IOCFR Link (CFR: 43.2 I45.2)

Page 119of 176

Millstone Unit 2 LOlT 2005 RO and SRO Exam Analysis (Questions with at least a 50% miss rate) 5 of 7 ROs missed

69 Unit 2 was operating at 100% power. Given the fobwing events and conditions:

- A reactor trip occurs at 0200 and the operators enter EOP-2525, Standard Post Trip Actions

- At 0210, several rad monitors are rising, including RM-4299 C (main steamline radiation monitor),

which slowly increases from 1.OE-2 to 1.5EO R/hr Which one of the following statements correctly describes the cause of this trend?

A A SGTR has occurred in the B S/G at 0210 1 6 A SGTR occurred on the 6 S/G prior to the trip at 0200 c A large crud burst has occurred in the RCS at 0200 Severe fuel damage has occurred at 0210 The candidates that chose A or B did so for 1 of 2 reasons: 1) Candidate did not remember that tlie Main Steam Line Rad Monitors would begin to read higher, post-trip, as a result of a fuel failure;

2) Candidate did not know specifically which additional radiation monitors were rising. The assumption was that only the Steam Jet Air E.jector and/or the Blowdown radiation monitors were rising, ivhich is indicative of a Steam Generator Tube Rupture. During the post exam review, the students were reminded that the Main Steam Liiie Radiation Monitors are NOT useful in determining the presence of a tube rupture after the trip due to the short half-life of N-16 gammas.

This question is technically accurate; however, it may be prudent to include more information on specifically which radiation monitors are also rising. For example, if the #2 Main Steam Line Radiation Monitor is rising due to failed fuel, then the # 1 Main Stearn Line Radiation Monitor would also be rising an equal or proportional amount. This indication would be very apparent to the operator in the Control Room. Additionally, other radiation monitors, easily accessible to the operator, would show a rise solely due to failed fuel. This question provides only one backup method to determine fuel failure, when other primary indications would be readily apparent to the operator.

If tlie question were modified to include the # I Main Steam Line Radiation Monitors, then distractors A and B would need to be modified to include a SGTR on A S/G as well as B S/G.

Page 10 of I 7

Millstone Unit 2 LOlT 2005 RO and SRO Exam Analysis (Questions with at least a 50% miss rate) 5 of 7 ROs missed: 2 of 2 SROs missed

71 A transfer of a new fuel assembly is in progress from one location in the spent fuel pool to another using OP-Z303B, SFP Fuel Handling Operations. The operator raises the hoist with the desired assembly grappled until upward motion is stopped by the upper limit switch interlock.

What must be done next?

7 A Release hoist raise switch, use the bridge/trolley controls to move to destination 0 Stop all hoist and crane movement and notify Reactor Engineering immediately c Lower assembly into initial location and contact Reactor Engineering for resolution.

D Slowly lower hoist until load cell indicates 250 to 290 pounds, then continue move Upon review. it was noted that both A and R are correct answers. (See attached justification.) It is rccommended that both A and B be accepted as correct and that this question be substituted

\\zit11 the attached to avoid future difficulties.

Pngc 1 1 o t 1 I

,/ RO ,/ SRO Question ID: 0053864 f 7 Memory7 (Check=Yes)

The plant IS In a refueling outage with fuel movement in progress The Refueling Machine IS currently performing a core to core move with an assembly in the Refueling Machine It has just been realized that the Refueling Machine mast is in the wrong position and needs to be rotated 90 degrees to protect the camera.

Which one of the following actions should be taken in this situation?

A Immediately stop fuel movement and have the Refueling Machine SRO direct subsequent Refueling Machine operations. Notify the Shift Manager when this evolution is complete.

B Rotate the mast 90 degrees as required, then complete this fuel movement, ensuring the bundle rotation is noted in the fuel movement log. Notify Reactor Engineering when this evolution is complete.

C Immediately stop fuel movement and get guidance from Reactor Engineering on necessary e actions. Have the Shift Manager or Refueling Machine SRO direct subsequent operations.

D Deposit the fuel assembly in an open core location closest to the desired one, rotate the mast, then re-grab the original fuel assembly and place it in the desired location. Notify Reactor Engineering when this evolution is complete.

Justification The fuel mast needs to be in the proper position to prevent damage to the TV camera, and the fuel assembly should not be rotated unless directed by Reactor engineering. Also, a fuel assembly can not be placed in any position not covered by the fuel movement procedure without specific permission by the RE.

Reference MP2'LORT94955 [034 REF-01-C49901 (12/9/97) OP 2393, FH NRC WA SystemlEIA NRC WA Generic System 2.2 Equipment Control Number G 2.2.28 Knowledge of new and spent fuel movement procedures.

Importance ROlSRO 2.6 3.5 IOCFR Link (CFR: 43.7 145.13)

Page 131 of 176

Millstone Unit 2 LOlT 2005 RO and SRO Exam Analysis (Questions with at least a 50% miss rate) 3 of 7 ROs missed

72 Refueling is in progress. A new fuel assembly has just been lowered into core location A-1 1 (core map attached). You are the PPO and have noted the following before and after readings on the wide range logarithmic power channels:

BEFORE AFTER WR CH A 1.9E1 cps 2.OE1 cps WR CH B 1.8E1 cps 3.2E1 cps WR CH C 1.6E1 cps 1.9E1 cps WR CH D 1.OE1 cps 1.2E1 cps Based on these indications, which of the following is required?

A Suspend all core alterations and positive reactivity additions 3 B Commence boration per AOP-2558, Emergency Boration Continue to monitor nuclear instruments, NO immediate action required D Withdraw the fuel assembly and contact Reactor Engineering for guidance This cl~iestiondid NOT have a greater than or equal to 50% fail rate; however, it was noted that the qucstion should be modified to provide only 1 correct answer versus the existing 2 correct answers.

Upon i-eiiew. it was noted that both C and D are correct answers. (See attached justification.) It is recommended that both C and D be accepted as correct and that this question be modified per the attached for future use.

Page 12 of 17

1. 72 1 v RO iv SRO Question ID: 5000066 1- 1 Memory? (Check=Yes)

Refueling is in progress A new fuel assembly has just been lowered into core location A-1 1 (core map attached) You are the PPO and have noted the following before and after readings on the wide range logarithmic power channels BEFORE AFTER WR CH A 1.9E1 cps 2.1E1 cps WR CH B 1.8E1 cps 2.3E1 cps WR CH C 1.6E1 cps 1.7E1 cps WR CH D 1.OE1 cps 1.OE1 cps Based on these indications, which of the following is required?

A Suspend all core alterations and positive reactivity additions.

B Commence boration per AOP-2558, "Emergency Boration".

C Continue to monitor nuclear instruments, NO immediate action required. Jl D Withdraw the fuel assembly and contact Reactor Engineering for guidance.

Justification CHOICE (A) - NO WRONG. Counts have not doubled. The only appreciable increase in counts is on CH B which is immediately adjacent to the location of the new assembly.

VALID DISTRACTOR: Per OP-2209A, if at any time, unanticipated count rate multiplication, (i e , doubling),

is indicated, then suspend refuel operations.

CHOICE (B) - NO WRONG: Counts have not doubled. The only appreciable increase in counts is on CH B which is immediately adjacent to the location of the new assembly.

VALID DISTRACTOR: Per OP-2209A, if at any time, unanticipated count rate multiplication, (i e., doubling),

is indicated. then commence boration CHOICE (C) - YES Criteria for action is observation of an unanticipated count rate multiplication, (Le., doubling). Counts have not doubled.

CHOICE (D) - NO WRONG: Counts have not doubled. The only appreciable increase in counts is on CH B which is immediately adjacent to the location of the new assembly.

VALID DISTRACTOR: Plausible that requirement is to remove the assembly to lower core reactivity while situation is evaluated.

Reference NRC-2005 NRC KIA System/E/A NRC KIA Generic System 2.2 Equipment Control 2.2 Equipment Control Number G 2.2.30 SEE GENERIC K/A "Knowledge of RO duties in the control room during fuel handling such as alarms from fuel handling area, communication with fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation."

Importance ROlSRO 3.5 3.3 IOCFR Link (CFR: 45.12)

Page 133 of 177

Millstone Unit 2 LOlT 2005 RO and SRO Exam Analysis (Questions with at least a 50% miss rate)

SRO ONLY; 1 of 2 SROs missed

77 The TS limiting condition for operation action requirements for a Safety Injection Tank (SIT) are less restrictive on time allowed to restore to OPERABLE status for boron concentration than for low level.

This is because the A tank volume requirements are based on minimizing the volume of nitrogen entering the SG tubes and preventing RCS heat removal.

B tank volume requirements are based on one tank emptying through the break and a passive failure of a second tank.

l a boron requirements consider the average concentration in the total volume of three safety injection tanks.

' D boron requirements assume sufficient shutdown margin due to void fraction during a large break LOCA.

This question is technically accurate. Knowledge deficiencies were covered during the post exam rc\ie\i.. No modification is required.

Page 13 of 17

Millstone Unit 2 LOlT 2005 RO and SRO Exam Analysis (Questions with at least a 50% miss rate)

SRO ONLY: 2 of 2 SROs missed

79 The following plant conditions exist:

The plant is at 100% power. The 'B' Emergency Diesel Generator (EDG) was declared INOPERABLE yesterday at 0600. At 0800 today, the Shift Manager discovers that the conditional 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months surveillance operability run on the ' A EDG, required by the 'B' EDG action statement has not been performed.

What action is required?

The operability surveillance of A EDG must be performed successfully by 1000 today or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

B The operability surveillance of A EDG must be performed successfully by 0900 today or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c The operability surveillance of A' EDG must be performed successfully by 1200 today or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

D The operability surveillance of A' EDG must be performed successfully by 0800 tomorrow or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

This question is technically accurate. Both candidates assumed that Tech Spec 3.0.3, which is more consenfati\e. applied for this condition. Knowledge deficiencies were covered during the post exam re\,ie\\,. N o modification is required.

Page 14 of 17

Millstone Unit 2 LOlT 2005 RO and SRO Exam Analysis (Questions with at least a 50% miss rate)

SRO ONLY; 1 of 2 SROs missed

82 Unit 2 is operating near beginning of cycle at a burnup of 600 MWD/MTU.

The following conditions exist AFTER a transient from 90% power:

- steam generator pressure is lower

- main generator megawatt output is lower

- indicated feedwater temperature is lower

- reactor coolant hot leg temperature is lower Which one of the following events caused this plant response and what is the applicable procedure for addressing the problem? Assume NO operator action.

A condenser backpressure rise (degraded vacuum), address with ARP-2590E (A-37), "COND VACUUM LO" B sensor input to throttle pressure limiter failed (0 psig), address with ARP-2590D (DA-22),

"10% TURBINE LOAD DECREASE" I C feedwater heater extraction steam isolation valve closed (heater 1B), address with ARP-25900 (AA-18), "HEATER 1A LEVEL HI" atmospheric dump stuck in intermediate posi!im (50% open), address with ARP-2590D (B-6),

"ATMOSPHERIC DUMP VALVE NOT CLOSED" This question is technically accurate. Knowledge deficiencies (Le., closure of extractioii steam valve does not lower generator megawatt output or steam generator pressure) were covered during the post e x a n review. No modification is required.

Page 15 of 17

Millstone Unit 2 LOlT 2005 RO and SRO Exam Analysis (Questions with at least a 50% miss rate)

SRO ONLY; 1 of 2 SROs missed

1. 90 Refueling IS in progress on Unit 2. During normal rounds, the Aux Building PEO reports that the red light on the SFP SW Area Radiation Monitor (RM-8139) local module is illuminated.

Which of the following is a possible reason for the reported indication?

A loss of power to the radiation monitor I B local horn silence switch in the OFF position actual high radiation condition in spent fuel pool area D Fuel Area Radn AEAS switch at ESF sensor cab in INHIBIT This question is technically accurate. Knowledge deficiencies (Le., the red light is NOT lit if the local horn silence switch is placed in OFF. unless a high radiation condition exists) were covered during the post e x m i review. No modification is required.

Page 16 of 17

Millstone Unit 2 LOlT 2005 RO and SRO Exam Analysis (Questions with at least a 50% miss rate)

SRO ONLY; 2 of 2 SROs missed

93 The unit is in a refueling outage. A modification was installed on the condenser air removal system to troubleshoot air in-leakage problems.

If this installation was performed as a "temporary modification", what time limitation is associated with the modification?

A Unless required more frequently by SORC, the modification shall be audited within 90 days after installation and at least once per calendar quarter thereafter.

B Unless required more frequently by SORC, the modification shall be audited within 90 days after installation and at least once per calendar year thereafter.

' c Unless authorized by Station Director, the modification shall be removed prior to the end of the next refuel outage or a time not to exceed 18 months, whichever is shorter.

Unless authorized by Station Director, the modification shall be removed prior to t h e end of the next refuel outage or a time not to exceed 24 months, whichever is shorter This question is technically accurate, Both candidares selected an answer concerning the uiiaximum length of time a temporary modification can remaitl in effect that was more conservative than the actual time limit. The knowledge deficiency was covered during the post exam review. No ni o d i fi cation i s required.

Page 17 Of' 17

Serial No. 04-57OF Docket No. 50-336 Attachment 2 March 2005 Reactor Operator and Senior Reactor Operator Technical Justifications Renardinq Questions 32.38,66,71 and 72 And Performance of JPM A2SRO SRO AWO Acceptance Millstone Power Station 2 Dominion Nuclear Connecticut, Inc. (DNC)

Question #32: NRC License Exam Key PDF file, page #35 (Attached)

References:

EOP-2532, Page 46 of 95, Step 54 (Attached)

EOP-2541, Appendix 18, Simultaneous Hot & Cold Leg Injection (Attached)

ECCS Training Figure 11A, Boron Precipitation 2 2 Not Available (Attached)

Comments:

The justification for the original correct choice B states that although it is physically possible to align A LPSI pump for Hot Leg Injection, there is no procedural guidance to do so. The reason there is no procedural guidance for this system alignment under the given conditions (24D deenergized), is because without 24D power it is not possible to align LPSI for Hot Leg Injection. One of the major valves in the flow path to achieve this alignment (2-SI-652) is powered by Facility Two (24D) and is located in containment (inaccessible with a LOCA). The procedure section the operators are referred to with 24D lost assumes Facility One HPSI & LPSI pumps are both available.

Therefore, as stated in Choice A, if the available HPSI is lost, core cooling would be in jeopardy because there is no system flowpath for Hot Leg Injection with the available LPSI pump.

Recommendation:

Based on the above explanation, we believe both Choice A and Choice B are technically correct for the given information in the stem.

I have reviewed the Comments and attached References for question #32 and concur with the

32 I 2RO PI SRO QuestionID: 5000029 Origin: New Memory Level The plant is operating at full power with all equipment functional, except for the 'B' HPSl Pump, which is 00s for maintenance.

Then, a large break LOCA occurs combined with a loss of Bus 24D (due to an electrical fault on 24D).

Which one of the choices correctly completes the following statement regarding the impact of the loss of ECCS pumps.

10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months after the event, a loss of the only available adversely affect long term core cooling because the remaining .

A HPSI pump would, LPSl pump does NOT have a system flowpath for boron precipitation 7 control.

B HPSl pump would, LPSl pump could NOT be procedurally realigned for boron precipitation d control via hot leg injection.

C LPSl pump would NOT, HPSI pump is preferred for boron precipitation control. I-1 D LPSl pump would NOT, HPSl pump could be procedurally realigned for boron precipitation 0 control via hot leg injection.

Justification

~~~

I CHOICE (A) - NO WRONG: LPSl does have a system flowpath for boron precipitation control.

VALID DISTRACTOR: Plausible that HPSl injection necessary for adequate injection flow.

CHOICE (B) - YES A single HPSl pump will provide sufficient flow for long term cooling. A LPSl pump could physically be aligned for hot leg injection but the EOPs do not provide procedural guidance for performing this task.

CHOICE (C) - NO WRONG: Loss of LPSl would have adverse affect because of inability to realign HPSI.

VALID DISTRACTOR: HPSl does provide sufficient core cooling flow.

CHOICE (D) - NO WRONG: HPSl could not be procedurally realigned for boron precipitation control because it is needed for injection.

VALID DISTRACTOR: HPSI could physically realigned but not iaw procedure.

References I

1. ECC-01-C. "Emergency Core Cooling System", Revision 3 (6/28/01) (Pg 1 I. I 3 of 25)

2. OP-2541, Appendix 18. "Simultaneous Hot and Cold Leg Injection" NRC KIA SystemlElA System 006 Emergency Core Cooling System (ECCS)

Number K6.13 RO 2.8 SRO 3.1 CFR Link (CFR: 41.7/45.7)

Knowledge of the effect of a loss or malfunction on the following will have on the ECCS: Pumps NRC KIA Generic System Number RO SRO CFR Link

Millstone Unit 2 EOP 2532 Revision 23 Loss of Coolant Accident Page 46 of 95 INSTRUCTIONS CONTINGENCY ACTIONS Hot and Cold Leg Injection

54. IF elapsed time from the start of the LOCA is between 8 to 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months AND A N Y of the following conditions exist:

CET subcooling is less than the minimum subcooling curve. Refer To Appendix 2, Figures.

Pressurizer level is less than 20%

Reactor vessel level is less than 43%

ESTABLISH simultaneous hot and cold leg injection. Refer To Appendix 18, Simultaneous Hot and Cold Leg Injection.

Operate Hydrogen Recombiners

55. IF containment hydrogen concentration is between 1.4 to 3.8%,

OPERATE hydrogen recombiners.

Refer To Appendix 20, Hydrogen Recombiner Operation.

Operate Hydrogen Purge System

56. IF containment hydrogen concentration I is greater than or equal to 2.9% AND the Technical Support Center concurs, OPERATE the Hydrogen Purge System.

Refer To Appendix 21, Hydrogen Purge System Operation.

R

MILLSTONE NUCLEAR POWER STATION EMERGENCY OPERATING PROCEDURE Simultaneous Hot and Cold Leg Injection EOP 2541, Appendix 18 Rev. 000 Approval Date: 10/2/03 Effective Date: 10/3/03

Millstone Unit 2 EOP 2541, Appendix 18 Revision 000 Simultaneous Hot and Cold Page 1 of 8 Leg; Injection INSTRUCTIONS CONTINGENCY ACTIONS IF BOTH Facility 1 and 2 are

-1.. -

available AND CET temperature is greater than 345"F, Go To Attachment 18-B.

-2.. -

IF Facility 2 power is available, Go To Attachment 18-A.

-3.. -

IF Facility 2 power is not available, Go To Attachment 18-B.

Millstone Unit 2 EOP 2541, Appendix 18 Revision 000 Simultaneous Hot and Cold Page 2 of 8 Leg Injection Appendix 18A Facility 2 Power Available Page 1 of 4 INSTRUCTIONS CONTINGENCY ACTIONS CAUTION v I Boron precipitation control steps require valve manipulations in a potentially high radiation area.

-1.. IF Facility 2 power is available,

~~

I INITIATE boron precipitation control via SDC suction by performing the following:

a. STOP BOTH LPSI pumps.

b. ENSURE Facility 2 HPSI b.1 Go To Attachment 18-B.

pump is operating.

c. E Facility 1 is energized, PERFORM the following:

1) CLOSE SI-635, LPSI 1) Go To Attachment 18-B.

injection valve.

2) ENSURE TWO of the following valves are fully closed and only ONE of the valves is fully open:

SI-615, LPSI injection valve SI-625, LPSI injection valve SI -645, LPSI injection valve

3) Go To step l..e.

(continue)

Millstone Unit 2 EOP 2541, Appendix 18 Revision 000 Simultaneous Hot and Cold Page u 3 of 8 Leg Injection Appendix 18A Facility 2 Power Available Page 2 of 4 INSTRUCTIONS CONTINGENCY ACTIONS

-1.. (continued)

NOTE With Facility 1 power unavailable? Electrical Maintenance may be contacted to determine SI-615 and SI-625 valve position using MCC test jacks at MCC B51.

d. IF Facility 1 power not available, PERFORM the following:

1) ALIGN SI-651, SDC suction isolation power to Facility 2 by peforming the following:

1.d.l)l. OPEN breaker B5110.

l.d.1)2. OPEN disconnect switch NB5110 (outside MCC B51) l.d.l)3. TURN and REMOVE kirk key from NB5110.

l.d.1)4. INSERT and TURN kirk key in NB6172.

l.d.l)5. CLOSE disconnect switch NB6172. (outside MCC B51) l.d.l)6. CLOSE disconnect switch NSI651. (outside MCC B51) l.d.1)7. CLOSE breaker B6172.

(continue)

Continuous

Millstone Unit 2 EOP 2541, Appendix 18 Revision 000 Simultaneous Hot and Cold Page 4 of 8 Leg Injection Appendix 18A Facility 2 Power Available Page 3 of 4 INSTRUCT1ONS CONTINGENCY ACTIONS NOTE With Facility 1 power unavailable, Electrical Maintenance may be contacted to determine SI-615 and SI-625 valve position using MCC test jacks at MCC B51.

-1.. (continued)

2) DETERMINE position of LPSI injection valves, SI-615 and SI-625.

(outside MCC B51)

3) IFLPSI injection valves SI-615 and SI-625 are deenergized open, CLOSE LPSI injection valves, SI-635 and SI-645. (CO1)

4) IF LPSI injection valves, SI-615 and SI-625 are deenergized closed, PERFORM the following:

l.d.4)l. OPEN LPSI injection valve 2-SI- 645.

1.d.4)2. CLOSE LPSI injection valve 2-SI- 635.

5) OPEN SI-651, SDC suction isolation valve.

(local panel C530)

(continue)

Millstone Unit 2 EOP 2541, Appendix 18 Revision 000 Simultaneous Hot and Cold Page 5 of 8 Leg Injection Appendix 18A Facility 2 Power Available Page 4 of 4 INSTRUCTIONS CONTINGENCY ACTIONS

-1.. (continued)

e. ENSURE TWO HPSI injection valves are fully open and TWO are fully closed:

(C01)

SI-616 SI-626 SI-636 SI-646

f. OPEN BOTH of the following: f.1 Go To Attachment 18-B.

SI-400, SDC warmup isolation valve (local)

SI-709, SDC suction isolation (local)

g. CLOSE 2-SI-652 Manual g.1 Go To Attachment 18-B.

Disconnect Switch, 89 -SI - 652. (West wall of Control Room)

h. IF SI-651 is not open, h.1 Go To Attachment 18-B.

CLOSE 2-SI-651, manual disconnect switch, NS651.

1. I F BOTH Facility 1 and

- i.1 Go To Attachment 18-B.

Facility 2 power available OPEN key locked SDC suction isolation valves, SI -65 1 and SI-652. (CO1)

J. IF only Facility 2 power j.1 Go To Attachment 18-B.

available, OPEN key locked SI -652, SDCS suction isolation valve.

(C01)

k. START ONE LPSI pump.

(CW Level of Use Continuous

Millstone Unit 2 EOP 2541, Appendix 18 Revision 000 Simultaneous Hot and Cold - 6 of 8 Page Leg Injection INSTRUCTIONS CONTINGENCY ACTIONS 8B Facility 2 Power NOT Available Page 1 of 3

-1.. IF Facility 1 power is available, INITIATE boron precipitation control via auxiliary pressurizer spray by the following:

a. E BOTH Facility 1 and 2 are available AND CET temperature is greater than 345 O F ,

PERFORM the following:

1) ENSURE Facility 2 HPSI pump is operating. (CO1)

2) ENSURE BOTH LPSI pumps are secured. (CO1)

3) Go To step l..e.

b. ENSURE Facility 1 HPSI pump is operating. (CO1) - ./

d v d , 4b / p pep $h?n h f o .

NOTE With Facility 2 power unavailable, Electrical Maintenance should be contacted to determine SI-635 and SI-645 valve position using MCC test jacks at MCC B61.

~ ~~

c. ENSURE TWO of the following valves are fully closed and TWO of the valves are fully open: (CO1)

SI-615 SI-625 SI -635 SI-645 (continue)

Level of Use

Millstone Unit 2 EOP 2541, Appendix 18 Revision 000 Simultaneous Hot and Cold Page 7 of 8 Leg Injection

d. ENSURE LPSI pump A operating and LPSI pump 3 is off.(C01)

e. STOP Facility 1 HPSI pump. A G A ~ :/4A b JF pfl ~:e ih

f. CLOSE Facility 1 HPSI injection valves, SI-617, 627, 637 and 647 (CO1).

g. UNLOCK and OPEN charging pump discharge to HPSI header valves, CH-340 and CH-440. (Boronometer room, 146 Aux Bldg)

h. CHECK open CH-429, charging header isolation valve. (C02)

1. OPEN CH-517, pressurizer i.1 IF Facility 2 DC power is not a m spray isolation. (C02) available, ALIGN CH-517 and CH-519 power to Facility 1 as follows:

1) OPEN breaker DV2012.

2) CLOSE breaker DV1012.

3) PLACE CS-DV20 2 2 POWER CH-517/CH-519 keylock switch to OFF.

(C02)

4) PLACE CS-DVlO Z1 POWER CH -5 17/CH-5 19 keylock switch to ON.

(C02)

5) OPEN CH-517, pressurizer aux spray isolation. (C02)

Level of Use

Millstone Unit 2 EOP 2541, Appendix 18 Revision 000 Simultaneous Hot and Cold Page 8 of 8 INSTRUCTIONS CONTINGENCY ACTIONS 8B Facility 2 Power NOT Available Page 3 of 3

j. CLOSE loop 1A and 2A charging isolations, CH-518 and CH-519. (C02)

Question #38: NRC License Exam Key PDF file, page #41 (Attached)

References:

AOP-2590D-025, Pages 1 & 2 (Attached)

Feedwater Control System Training Material, Pages 32 & 33 (Attached)

Comments:

Our original concern with this question was that it required the Candidate to go beyond the knowledge solicited by the WA (i.e.; how will the system respond to the failed instrument), and make a quantitative judgment as to the amount the system will respond to the failed instrument. That was the reason behind our suggested rewrite of the question (see attached) to eliminate the choice of . ..low level reactor trip. and replace this distracter with one that was clearly wrong in its magnitude of response. In disallowing this suggested change, the Candidate was forced into a quantitative judgment, which depending on the specific tuning of the system by I&C for that fuel cycle, could result in either Choice A or Choice B being the correct response.

Recommendation:

Based on the above explanation, we believe both Choice A and Choice B are acceptable answers for the given information in the stem.

I have reviewed the Comments and attached References for question #38 and concur with the

38 I ERO 1 9 1SRO QuestionID: 5000034 Origin: Mod I 1 Memory Level Given the following plant conditions

- 100% power

- SG levels at setpoint

- Steam flow and feed flow matched

- SG2 Feed Flow Transmitter FT-5269A output fails high With NO operator actions, which of the following describes the expected plant response?

A SG level lowers, but stabilizes above the low level reactor trip. 4 B SG level lowers to the low level reactor trip. 1 C SG level rises, but stabilizes below the high level turbine trip. (J D SG level rises to the high level turbine trip. ' I Justification CHOICE (A) - YES The relatively small change in the magnitude of the averaged feed flow signal with this failure occurring at 100% power is not sufficient to drive level to the low level trip. The controller will a d on the level deviation signal to restore level to setpoint.

CHOICE (B)- NO WRONG: Output from feed flow transmitters FT-5269A and FT-5269B on the SG2 feed line are averaged for input to the threeelement level control. Failing one transmitter high drives the average high. However, at full power, the resulting charge in the magnitude of the averaged signal is small. The control system will respond by throttling closed on the FRV. The level signal MI1 act on the steam flow signal to moderate the response, restoring level to setpoint.

VALID DISTRACTOR: Applicant may think that feed flow/steam flow mismatch will drive level to the low level trip setpoint.

CHOICE (C) - NO WRONG: SG will lower to the low level trip setpoint.

VALID DISTRACTOR: Applicant may think the higher indicated feed flow will cause SG level to rise but stabilize below the high level trip based on input to the control system from the level signal.

CHOICE (D) - NO WRONG: SG will lower to the low level trip setpoint.

VALID DISTRACTOR: Applicant may think the higher indicated feed flow will cause SG level to rise to the turbine trip.

References I

1. FWC-01-C, "Feedwater Control System". Revision 2 (3/22/04) (Pg 7,8 of 46)

2. Source: INPO Bank - # 1942 - Used at Palisades 1,6/14/1999 NRC KIA SystemlElA System 059 Main Feedwater (MFW) System Number K4.08 RO 2.5 SRO 2.7 CFR Llnk (CFR: 41.7)

Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following: Feedwater regulatory valve operation (on basis of steam flow, feed flow mismatch)

NRC KIA Generic System Number RO SRO CFR Llnk

38 1 p~RO SRO Question ID. 5000034 I Origin: Mod I 0 Memory?(Check=Yes)~

Given the following plant conditions: p,q- Kxdm $ ~ g j e sbed d h ~ ~ f e

- 100% power

- SG levels at setpoint

- Steam flow and feed flow matched

- SG2 Feed Flow Transmitter FT-5269A output fails high With NO operator actions, which of the following describes the expected plant response?

A SG level will lower when the Feed Reg. Valve fully closes on the flow mismatch. 0 B SG level will lower on the flow mismatch, but stabilize above the low level RPS trip.

c SG level will rise on the feed flow dominant signal until the high level turbine trip. 0 D SG level will rise on the feed flow demand, but stabilize below the high level turbine trip. 0 Justification CHOICE (A) - NO WRONG: Output from feed flow transmitters FT-5269A and FT-5269B on the SG2 feed line are averaged for input to the three-element level control. Failing one transmitter high drives the average high. The control system will respond by closing the FRV. The level signal will slowly act on the steam flow signal to moderate the response. However, because of the relatively small enor of the feed flow signal and the fact that the signal is STEAM flow dominant, will result in little effect on the level input signal.

VALID DISTRACTOR: Applicant may think that feed flow signal will dominate level control, resulting in SG level continuing to lower.

CHOICE (B) YES -

CORRECT: Output from feed flow transmitters FT-5269A and FT-5269B on the SG2 feed line are averaged for input to the three-element level control. Failing one feed flow transmitter high drives the average up a small amount. The control system will respond by closing the FRV. The level signal will slowly act on the steam flow signal to moderate the response. Without operator action, level will decrease a small amount, but soon be tumed by both level error and steam flow dominance.

CHOICE (C) - NO WRONG: SG will lower to the low level trip setpoint.

VALID DISTRACTOR: Applicant may think that the feed flow error dominates the level control system, causing SG level to rise to the turbine trip.

CHOICE (D) - NO WRONG: SG will lower to the low level trip setpoint.

VALID DISTRACTOR: Applicant may think the higher indicated feed flow will cause SG level to rise but stabilize below the high level trip based on input to the control system from the level and steam flow signal.

Reference (059 FWC-01-C R 0 9 8858), NRC-2005 Question Original question was not technically correct for design of FWLCS.

Hlstory NRC K/A System/E/A NRC WA Generic System 059 Main Feedwater (MFW) System Number K4.08 Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following:

Feedwater regulatory valve operation (on basis of steam flow, feed flow mismatch)

Importance ROISRO 2.5 2.7 10CFR Link (CFR: 41.7)

Page 42 of 109

01/21/04 02/12/04 Approval Date Effective Date Setpoint: 80% - High 60% - LOW 1 A-7 1 STEAM GENERATOR 1 LEVEL HI/LO AUTOMATIC FUNCTIONS

1. None CORRECTIVE ACTIONS

1. OBSERVE No. 1 SG level and DETERMINE actual level (C-05 , PPC).

CAUTION v

1. High SG levels, (SG overfill), may cause carryover and subsequently cause main turbine or main steam line damage. If No. 1SG level reaches 85%, No. 1 FRV and FRV bypass valve ramp closed followed by FRV lockup.

2. If SG level is low, to ensure feed rate does not cause reactor ower transients reaching reactor trip set oints or cause level shrinkage, P

caution must be used in the rate o opening FRV when restoring level.

2. Refer To OP 2385, Feedwater Control System Operation and CONTROL No. 1 FRV or FRV bypass valve in Manual (C-05).

3. RESTORE No. 1 SG level to normal band between 60 and 75%.

4. As required, ADJUST SGFP(s) speed to maintain FRV D/P between 20 and 100 psid (C-05).

5. IF No. 1 FRV does not respond to Manual control, MAINTAIN No. 1 SG level using SGFP(s) speed (C-05).

6. IF necessary, Refer To OP 2385, Feedwater Control System Operation and PLACE No. 1 FRV in Local Manual Operation.

7. DETERMINE cause of abnormal No. 1 SG level and CORRECT.

8. As desired, Go To OP 2385, Feedwater Control System Operation and RETURN No. 1 FRV and FRV bypass to Automatic.

ARP 2590D -025 Rev. 000 Page 1 of 2

SUPPORTING INFORMATION

1. Initiating Devices e High - FA-5268-10 e LOW- FA-5268-11

2. Computer Points L5272

3. Procedures e OP 2385, Feedwater Control System Operation

4. Control Room Drawings e 25203-26005, Sh.2 e 25203-32012, Sh.31

5. Annunciator Card Location: TB15-Jl3 ARP 2590D -025 Rev. 000 Page 2 of 2

Lesson

Title:

Feedwater Control System Page 32 of 46 Revision: 2 ID Number: FWC-01-C TEXT ACTIVITIES/NOTES E. MALFUNCTIONS AND FAILURES

1. Loss of Control Power/Air to FRVs When a loss of control power or a loss of instrument air to the FRV RO-8A positioner occurs an alarm is annunciated on (2-05. This alarm, Figure 5 FEEDWATER REGULATING VALVE l(2) LOCKED, alerts the control room operator to look at the white indicating lights on the bench section of C-05 to identify what caused the lock up.

At this time the FRVs are locked as is, so if possible, plant power RO-8B should be maintained as constant as possible. When the condition that caused the problem has been corrected and air or power has been returned, the control room operator can re-gain control of the FRVs by pressing the DOWNCOMER RESET pushbutton(s).

Pressing this button re-energizes the solenoid air valves which allows the FRVs to be positioned by the SGWLCS output signal.

2. SG Level Detector Failure Normally, the SG level AP transmitter switch is selected to BOTH. In RO-9A this position the SGWLCS selects the lowest of the two transmitter Figure 1 signal inputs. If one of the transmitters fail high this would have no effect on the SGWLCS output signal. Total transmitter failure would be indicated by the GREEN indicating light above the switch going out. If a failed high level transmitter is inadvertently selected and the SG HI LVL TURBINE TRIP permissive switch is in PERMISSIVE a Main Turbine trip will occur. If above 15% power a concurrent reactor trip will also occur.

If one of the SG level transmitters were to fail low, the FRV would RO-9B ramp open and try to restore indicated level to setpoint. The operator would need to take manual control of the FRV. Then the SG level AP transmitter switch is shifted to the operable transmitter. After establishing level at setpoint the operator can then shift SGW LCS back to automatic.

3. Feed Flow Detector Failure Normally, the feed flow AP transmitter switches are selected to 30-9A BOTH. In this position, the Spec 200 averages the two input signals

=igure 1 for its output flow signal. Depending on how the AP transmitter fails, high or low, will cause the output signal (average value) to read high or low. The SGWLCS will see a high or low mismatch between the steam flow and feed flow due to the faulty output signal from the feed flow circuit. The SGWLCS will position the FRV corresponding to the mismatch.

If this were to occur the operator would need to take manual control 30-96

Lesson

Title:

Feedwater Control System Page 33 of 46 Revision: 2 ID Number: FWC-01-C TEXT ACT IVITIES/NOTES of the FRV. Then the feed flow AP transmitter switch is shifted to the operable transmitter, as determined by the control room operator. After establishing SG level at setpoint the operator can then shift the FRV back to automatic.

4. Steam Flow Detector Failure Normally, the steam flow AP transmitter switches are selected to RO-9A BOTH. In this position, the Spec 200 averages the two input signals Figure 1 for its output flow signal. Depending on how the AP transmitter fails, high or low, will cause the output signal (average value) to read high or low. The SGWLCS will see a high or low mismatch between the steam flow and feed flow due to the faulty output signal from the steam flow circuit. The SGWLCS will position the FRV corresponding to the mismatch. In addition, the SGFP speed control system will receive the faulty output signal from Spec 200. This faulty signal will cause BOTH SGFPs to either raise or lower their speed depending on how the AP transmitter fails. This change in SGFP speed will also change feed flow which will be seen by the SGWLC for both FRVs.

If this were to occur, the operator would need to take manual control of one or both of the FRVs and BOTH SGFPs to stabilize the plant.

After stabilizing the plant, the steam flow AP transmitter switch is shifted to the operable transmitter, as determined by the control room operator. After establishing SG level at setpoint the operator can then shift the FRV and both SGFPs back to automatic.

F. OPERATING EXPERIENCES

1. SER 1-83 Rapid, Single Loop Cooldown Transient From Excessive Feedwater After Trip During a plant startup at 20 percent power, feedwater (10 percent of total rated flow to each of the two steam generators) was being supplied primarily through the bypass valves rather than the main feedwater control valves. An instrument technician opened the door to the feedwater control panel which inadvertently unplugged the power supply to the panel. By design, the main feedwater control valves failed-as- is, and the bypass valves failed shut.

The reduced feedwater flow resulted in decreasing steam generator level. In response, the operator attempted to increase flow by opening one of the two main control valves using remote manual control. However, the loss of feedwater control power prevented opening the valves by either automatic or remote manual means. The steam generator levels continued to drop, and the reactor operator manually tripped the reactor. Another technician, in responding to an annunciator for instrument bus power failure, found the feedwater control panel de-energized. He

Question #66: Attached

References:

Attached UNIT 2 TECHNICAL SPECIFICATIONS

1. TS 3.3.3.1 Table 3.3-1 Reactor Protective Instrumentation for Item 11 and associated ACTION 4

2. TS 3.1.1.1 SHUTDOWN MARGIN

3. Limiting Condition for Operation 3.0.4

4. BASIS for Specification 3.0.4 OP 2202 Reactor Startup IPTE Comments:

Choice D (the correct answer) is directly related to ACTION 4 for verifying compliance with TS 3.1.1.1 SHUTDOWN MARGIN, which is applicable for the current mode of operation (MODE 3). Choice D is an acceptable answer but is not a complete answer, taken by it self, because it implies that startup may continue.

Choice B Insert all control rods and shutdown the reactor is also an acceptable response because it is a proceduralized action of OP 2202 Reactor Startup IPTE and is the conservative philosophy of operations in DNAP- 14 10 Reactivity Management.

These actions prevent non-compliance with TS LCO 3.0.4 and its BASIS, which requires exercise of good practice in restoring systems or components to OPERABLE status before plant startup.

The Examinee may have been more compelled to place emphasis on compliance with reactor startup termination than shutdown margin verification because the startup procedure had just previously verified adequate SDM and has controls in place to continuously monitor conditions that could affect SDM (e.g.. ECP). It would be unacceptable to maintain current plant conditions considering the time required to re-verify SDM by obtaining and analyzing an RCS boron sample.

Recommendation:

Accept Choice D and Choice B I have reviewed the Comments and attached References for question #66 and concur with the recommendation to accept both Choice B and Choice D as correct answers.

/

Justification:

Choice B correctly describes the procedurally required termination of the reactor startup that is implicit in compliance with TS LCO 3.0.4 and its BASIS.

TS LCO 3.0.4 precludes Entry into an OPERATIONAL MODE or other specified condition when the conditions for the Limiting Condition for Operation are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval. Its BASIS goes on to say that the provisions of this specification should not, however, be interpreted as endorsing the failure to exercise good practice in restoring systems or components to OPERABLE status before plant startup.

DNAP-1410 Reactivity Management If the start-up is suspended near the point of criticality for an extended period of time, without a clear success path, then the core shall be made sufficiently subcritical to meet applicable shutdown margins.

OP 2202 Reactor Startup IPTE Prerequisite 2.1.15 requires a minimum of 2 Wide Range Logrithmic Monitors to be operable.

Imbedded throughout the procedure are requirements for continuously monitoring nuclear instruments, performance of channel checks and conservative response to anomalous behavior (Precaution 3.6 i.e. CEA Insertion).

In addition, the performance of the 1/M requires the use of at least 2 wide range NI drawers. This 1 / M is used in part to prevent criticality outside of a specified band. This band is more conservative than that required by TS 3.1.1.2 REACTIVITY BALANCE.

Due to the lack of adequate monitoring capability and an inability to accurately predict criticality it would be appropriately conservative for the operator to take these conditional actions:

3. IF at any time during reactor startup, it appears that criticality is reached, or is predicted to be reached, outside plus or minus 0.5% Delta rho (0.9% Delta rho for initial startup after refueling) band of ECP, PERFORM the following:

3.1 INSERT all CEA regulating groups in sequence (C-04).

3.2 REQUEST Chemistry Department sample and determine RCS boron concentration.

3.3 INITIATE a CR for Reactivity Management tracking.

3.4 Refer To OP 2208, Reactivity Calculations and, independent of CEA position, ENSURE adequate SHUTDOWN MARGIN using OPS Form 2208-013, Shutdown Margin Determination.

3.5 NOTIFY Reactor Engineering.

1. 66 ] q RO VI SRO QuestionID: 5000069 1 -1 ~1 ~

Memory?(Check=Yes)

Unit 2 IS conducting a reactor start up. Given the following events and conditions

-Wide range (WR) logarithmic nuclear instrument (NI) channels C and D are out of service

- The reactor is not yet critical

- The ECP expected critical rod height is 100 steps on Regulating Group 6

- Regulating Group 4 is withdrawn to 60 steps

- WR NI Channel A failed low WRL NI Channel A <l.OE-I CPS WRL NI Channel B 6.2E2 CPS Which one of the following statements correctly describes the required action (if any) required to comply with Technical Specifications?

A Immediately trip the reactor.

B Insert all control rods and shutdown the reactor.

C Stop the startup until WRL NI Channel A has been repaired. NO other actions are required D Immediately ensure adequate shutdown margin. V' Justification Tech Spec 3.3.1 requires 2 channels of WR Nl's to be operabte in MODES 3, 4 and 5. The reactor does not reach MODE 2 until group 4 CEAs are withdrawn t o & ?steps. Tech Spec LCO 3.3.1.1 applies under these conditions and requires immediate determination of adequate shutdown margin. There are no other Tech Spec requirements that apply to this case.

CHOICE [A] - NO WRONG Tech specs require 2 channels of WRL Nl's to be oprable in MODE 3. Tripping the reactor is not required - only determining adequate shutdown margin.

CHOICE [B] - NO WRONG Tech specs require 2 channels of WRL Nl's to be operable in mode 3. Immediately determining shutdown margin is required with the reactor in modes 3.

CHOICE [C] - NO WRONG -Tech Spec 3.3.1.1 requires 2 WRL NI channels to be operable in mode 3. Stopping the startup may be a prudent action but it is not required by Tech Spec 3.3.1 1.

CHOICE [D] - YES CORRECT The reactor has not yet transitioned to mode 2 and 2 WRL NI channels are required in mode 3 With only 1 WRL NI channel operable, the tech spec action is to immediately determine shutdown margin.

Reference NRC-2005 NRC KIA SystemlEIA NRC WA Generic System 033 Loss of Intermediate Range 2.2 Equipment Control Nuclear Instrumentation Number GA 2.2.22 SEE GENERIC WA Knowledge of limiting conditions for operations and safety limits Importance ROlSRO 3.4 4.1 10CFR Link (CFR 43 2 I45.2)

Page 118 of $76

MILLSTONE POWER STATION GENERAL OPERATING PROCEDURE Reactor Startup IPTE OP 2202 Rev. 020-06 Approval Date: 12/6/04 Effective Date: 12/16/04

. _. - .-~-~~~~ _ _ _ _ _ _ _ ~- ~

Attachment 5 Reactor Startup Conditional Actions (Sheet 1 of 3)

1. E at any time, the following conditions occur, PERFORM the specified action:

IF Tavglowers to between 515 and 525 O F AND the reactor is critical, Refer to OPS Form 2619A-001, Control Room Daily Surveillance and RECORD RCS temperature once every hour.

IF Tav lowers to less than 515 O F AND the reactor is critical, PERFORM the fohowing:

0 RAISE T& to greater than 515 O F within 15 minutes.

0 ETavgis riot greater than 515 O F within 15 minutes, PLACE plant in HOT STANDBY condition within the next 15 minutes.

0 Refer To TIS LCO 3.1.1.5 and DETERMINE applicability.

IF an uncontrolled cooldown occurs (Tc less than 500 OF), PERFORM the following:

0 TRIP reactor and INITIATE EOP 2525, Standard Post Trip Actions.

0 STOP one of the 4 operating RCPs (C-04).

0 Refer To AOP 2558, Emergency Boration, and INITIATE emergency boration.

0 Refer To TIS LCO 3.4.9.1 and DETERMINE applicability.

2. IF at any time a sustained SUR of 1.0 dpm is achieved, TRIP reactor and Go To EOP 2525, Standard Post Trip Actions.

4 Epredicted at any time during reactor startup, it appears that criticality is reached, or is to be reached, outside plus or minus 0.5% A e (0.9% A@for initial startup after refueling) band of ECP, PERFORM the following:

3.1 INSERT all CEA regulating groups in sequence (C-04).

3.2 REQUEST Chemistry Department sample and determine RCS boron concentration.

I 3.3 INITIATE a CR for Reactivity Management tracking.

3.4 Refer To OP 2208, Reactivity Calculations and, independent of CEA position, ENSURE adequate SHUTDOWN MARGIN using OPS Form 2208-013, Shutdown Margin Determination.

I 3.5 NOTIFY Reactor Engineering.

OP 2202 Sol T tirlvK ACT HCVICW Rev. 020-06

, Continuous 37 of 47

z c.

r t;

4 0

2:

n C

I Z FUNCTIONAL UNIT I

TOTAL NO.

OF CHANNELS REACTOR PROTECTIVE INSTRUMENTATION MINIMUM CHANNELS OPERABLE ACTION 1

N

1. Manual Reactor Trip 2 1

2. Power Level - High 3 2

3. Reactor Coolant Flow - Low 3 2

4.

w Pressurizer Pressure - High 3 2 h)

5. Containment Pressure - High 3 2

6. Steam Generator Pressure - Low 3 2

7. Steam Generator Water 3 2 Level - Low

8. Local Power Density - High 3 2 3 9. Thermal MargidLow Pressure 3 2 3

CD 10. Loss of Turbine - Hyraulic 3 2 52 CD Fluid Pressure - Low z

0 g

gN N

VI N

0 00 0 E

TABLE 3.3- 1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 8 Wide Range Logarithmic Neutron 4 0 2 3, 4, 5 rn Flux Monitor - Shutdown

12. DELETED

13. Reactor Protection System Logic 6 1 6 1 , 2 and

5 Matrices

14. Reactor Protection System Logic 4/Matr ix 3lMatrix 4lMatrix 1 , 2 and

6 Matrix Relays

15. Reactor Trip Breakers 4 3 4 1 , 2 and

6 z

?

TABLE 3.3-1 (Continued)

TABLE NOTATION

With the protective system trip breakers i n thc closed position and thc CEA drivc systcm capablc of' CEA withdrabval.

(a) Trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall bc automatically removed when THERMAL POWER is 2 5% of RATED THERMAL POWER (b) Trip may be inanually bypassed when steam generator pressure is < 800 psia and all CEAs are fully inserted; bypass shall be automatically removed when steam gcncrator pressurc is 2 800 psia.

(c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall bc automatically removed when THERMAL POWER is 2 1.5% of RATED THERMAL POWER.

(d) Trip does not need to be operable if all the control rod drive mechanisms are de-energized or if the RCS boron concentration is greater than or equal to the reheling concentration of Specification 3.9. I .

(c) DELETED

( f ) AT Power input to trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 2 5% of RATED THERMAL POWER.

ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and/or open the protective system trip breakers.

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may continue provided the following conditions are satisficd:

a. The inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The inoperable channel shall either be rcstorcd to OPERABLE status, or placed in the tripped condition, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

b. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , all functional units receiving an input from the inoperable channel are also declared inoperable, and the appropriate actions are takcn for the affected functional units.

c. The Minimum Channels OPERABLE requirement is met; howevcr, onc additional channel may bc removed from service for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , provided one of thc inoperable channels is placed in the tripped condition.

MILLSTONE - UNIT 2 314 3-4 Amendment No. 9,#, 32, H-6, -W,

%,226,280

Scptcmber 25. 2003 TABLE 3.3- I (Continued)

ACTION STATEMENTS ACTION 3 - NOT USED With the number of channels OPERABLE one less than rcquircd bjr the Minimum Channels OPERABLE requirenicnt, immediately verify compliancc with thc SHUTDOWN MARGIN requirements of Specification 3.1.1. I or 3.1.1.2, as applicable, and at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months thereafter.

ACTION 5 - With the numbcr of channels OPERABLE one less than requircd by the Minimum Channels OPERABLE requirement, restore the inoperablc channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within thc next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 6 - With the number ofchannels OPERABLE one less than requircd by the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MILLSTONE - UNIT 2 314 3-5 Amcndment No. 225. 282

September 25, 2003

-3 4 1 REACTIVITY CONTROL SYSTEMS 3 4.1.1 REACTIVITY CONTROL SYSTEMS

~~

SHUTDOWN MARGIN - (SDM)

LIMITING CONDITION FOR OPERATION The SHUTDOWN MARGIN shall be within the limit specified in the CORE OPERATING LIMITS REPORT.

APPLICABILITY: MODES 3 ( ] ) ,4 and 5 .

ACTION:

With the SHUTDOWN MARGIN not within the limit specified in the CORE OPERATING LIMITS REPORT, within 15 minutes, initiate and continue boration at 2 40 gpm of boric acid solution at or greater than the required refieling water storage tank (RWST) concentration (ppm) until the SHUTDOWN MARGIN is restored to within limit.

SURVEILLANCE REOUIREMENTS 4.1.1.1 Verify SHUTDOWN MARGIN is within the limit specified in the CORE OPERATING LIMITS REPORT at least once everv 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(I) See Special Test Exception 3 . I O . 1 MILLSTONE - UNIT 2 314 1-1 Amendment No. 33,6-k, *,a, 439, 448,280

f September 14, 2000 ->

3.4 LIMITING CONDITIONS FOR OPERATlON AND SURVEILLANCE REQUIREMENTS 3.,'4.0

..- APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals, except as provided in LCO 3.0.6. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour ACTION shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:

1. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time it is identified that a Limiting Condition for Operation is not met. Exceptions to these requirements are stated in the individual specifications.

This specification is not applicable in MODES 5 or 6. I Entry into an OPERATIONAL MODE or other specified condition shall not be made when the conditions for the Limiting Condition for Operation are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL MODE or specified condition may be made in accordance with ACTION requirements when conformance to them permits continued operation of the facility for an unlimited period of time. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements.

3.0.5 When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: ( I ) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system(s),

subsystem(s), train(s), component(s) and device(s) are OPERABLE, or likewise satisfy the requirements of this specification. Unless both conditions ( I ) and (2) are satisfied within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , ACTION shall be initiated to place the unit in a MODE in which the applicable Limiting Condition for Operation does not apply by placing it, as applicable, in:

MILLSTONE - UNIT 2 3/4 0-1 Amendment Nos. 62, W ,238,249

February 26, 199 1 APPLICABILITY The time limits of Specification 3.0.3 allow 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months for the plant to be in the COLD SHUTDOWN MODE when a shutdown is required during the POWER MODE of operation. If the plant is in a lower MODE of operation when a shutdown is required, the time limit for reaching the next lower MODE of operation applies. However, if a lower MODE of operation is reached in less time than allowed, the total allowance time to reach COLD SHUTDOWN, or other applicable MODE, is not reduced. For example, if HOT STANDBY is reached in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the time allowed to reach HOT SHUTDOWN is the next 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months because the total time to reach HOT SHUTDOWN is not reduced from the allowable limit of 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months . Therefore, if remedial measures are completed that would permit a return to POWER operation, a penalty is not incurred by having to reach a lower MODE of operation in less than the total time allowed.

The same principle applies with regard to the allowable outage time limits of the ACTION requirements, if compliance with the ACTION requirements for one specification results in entry into a MODE or condition of operation for another specification in which the requirements of the Limiting Condition for Operation are not met. If the new specification becomes applicable in less time than specified, the difference may be added to the allowable outage time limits of the second specification. However, the allowable outage time limits of ACTION requirements for a higher MODE of operation may not be used to extend the allowable outage time that is applicable when a Limiting Condition for Operation is not met in a lower MODE of operation.

The shutdown requirements of Specification 3.0.3 do not apply in MODES 5 and 6, because the ACTION requirements of individual specifications define the remedial measures to be taken.

blishes limitations on MODE changes when a Limiting Condition for peration 1s not met. It precludes placing the facility in a higher MODE of operation when the requirements for a Limiting Condition for Operation are not met and continued noncompliance to these conditions would result in a shutdown to comply with the ACTION requirements if a change in MODES were permitted. The purpose of this specification is to ensure that facility operation is not initiated or that higher MODES of operation are not entered when corrective action is being taken to obtain compliance with a specification by restoring equipment to OPERABLE status or parameters to specified limits. Compliance with ACTION requirements that permit continued operation of the facility for an unlimited period of time provides an acceptable level of safety for continued operation without regard to the status of the plant before or after a MODE change. Therefore, in this case, entry into an OPERATIONAL MODE or other specified condition may be made in accordance with the provision of the ACTION requirements.

The provisions of this specification should not, however, be interpreted as endorsing the failure to exercise good practice in restoring systems or components to OPERABLE status before plant startup.

MILLSTONE - UNIT 2 B 314 0-3 Amendment Nos. 62, 15 1

A P P L1CAB 1L ITY BASES (Con't)

When a shutdown is required to comply with ACTION requirements, the pro\isions of Specification 3.0.4 do not apply because they would delay placing the facility in a lower MODE of operation.

Speci tication 3.0.5 delineates what additional conditions must be satisfied to pennit operation to continue, consistent with the ACTION statements for power sources, when a normal or cmeigency power source in not OPERABLE. It specifically prohibits operation when one di\kion is inoperable because its norinal or emergency power source is inoperable and a system, subsystem, train, component or device in another division is inoperable for another reason.

The provisions of this specification permit the ACTION statements associated with individual systems, subsystems, trains, components, or devices to be consistent with the ACTION statements of the associated electrical power source. It allows operation to be governed by the time limits of the ACTION statement associated with the Limiting Condition for Operation for the normal or emergency power source, not the individual ACTION statements for each system, subsystem. train, component or device that is determined to be inoperable solely because of the inoperability of its nonnal emergency power source.

For example, Specification 3.8. I . I requires in part that two emergency diesel generators be OPERABLE. The ACTION statement provides for a 72-hour out-of-service time when one einergency diesel generator is not OPERABLE. If the definition of OPERABLE were applied without consideration of Specification 3.0.5, all systems, subsystems, trains, components and devices supplied by the inoperable emergency power source would also be inoperable. This would dictate invoking the applicable ACTION statement for each of the applicable Limiting Conditions for Operation. However, the provisions of Specification 3.0.5 perinit the time limits for continued operation to be consistent with the ACTION statement for the inoperable emergency diesel generator instead, provided the other specified conditions are satisfied. In this case, this would mean that the corresponding nonnal power source must be OPERABLE, and all I-edundant systems, subsystems, trains, components, and devices must be OPERABLE, or otherwise satisfy Specification 3.0.5 (i.e., be capable of performing their design function and have at least one normal or one emergency power source OPERABLE). If they are not satisfied, ACTION is required in accordance with this specification. I As a further example, Specification 3.8.1.1 requires in part that two physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system be OPERABLE. The ACTION statement provides a 24-hour out-of-service time when both required offsite circuits are not OPERABLE. If the definition of OPERABLE were applied without consideration of Specification 3.0.5, all systems, subsystems, trains, components and devices supplied by the inoperable normal power sources, both of the offsite circuits, would also be inoperable. This would dictate invoking the applicable ACTION statements for each of the applicable LCOs. However, the provisions of Specification 3.0.5 perinit the time limits for continued operation to MILLSTONE - UNIT 2 B 314 0-4 Amendment Nos. 78, W, 04-MP2-0 1G

MILLSTONE POWER STATION GENERAL OPERATING PROCEDURE all Reactor Startup IPTE OP 2202 Rev. 020-06 Approval Date: 12/6/04 Effective Date: 12/16/04

1. PURPOSE 1.1 Objective This procedure provides instructions for plant operations from OPERATIONAL MODE 3 (HOT STANDBY) to OPERATIONAL MODE 2 (STARTUP).

1.2 Discussion A reactor startup is a tremendous reactivity management concern. As such, conservative action shall be a part of all actions during the entire startup. The Control Room primary focus will be upon the reactor and its support systems, not on Work Control or other peripheral concerns. If at anytime during this startup the condition of the reactor and its responses are not understood, and controlled by the operators, the reactor must immediately be tripped and the actions of EOP 2525, Standard Post Trip Actions performed.

This procedure will ensure that all required Technical Specifications for entering Mode 2 will be met. Mode 2 will be entered when Group 4 regulating rods are at 72 steps.

Steps in this procedure may be performed in parallel or out of sequence, provided the SM or US reviews the applicable steps and determines that no plant conditions or system alignments established by any preceding steps are required, prior to commencing these actions.

Except where specified, the Shift Manager or Unit Supervisor must initial and date those steps that apply and N/A those steps that do not apply, as well as document reasons in Section 5, except for those instances where choices are given.

The approach to criticality must be carried out in a very careful and precise manner. The accepted method is to add reactivity in incremental steps, SUR is administratively limited to 0.5 DPM. The Operator stops after each incremental step to check for proper instrument response and to determine the state of the reactor. CEA withdrawal is stopped and Inverse Count Rate Ratio calculations ( l / ~ performed

) when count rate doubling occurs. The I@

Reactor Operator and Reactor Engineer are concurrently responsible for observing and declaring, to the Control Room staff, each count rate doubling.

ECP calculations should be performed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of expected criticality, and the expected time of criticality should be within 1hour of the estimated critical time used for ECP calculation unless performing a xenon-free startup. The reactor is declared critical when power is rising (i.e. positive SUR) with no CEA motion. Power is then raised to approximately lx10e3%

and critical data is recorded.

OP 2202 Continuous R Rev. 020-06 2 of 47

2.1.9 Audible count rate monitoring is available in Control Room.

2.1.10 Attachment 2, Requirements for Entering OPERATIONAL MODE 2 has been completed.

2.1.11 Reactor Engineering is available in the Control Room to perform the following functions:

An independent ECP ECP evaluations 1/M calculations and plotting Observing and declaring, to the Control Room staff, each count rate doubling.

2.1.12 The PPC is in service and operating with no computer testing in progress.

2.1.13 Health Physics Department has been notified of the impending reactor startup.

2.1.14 Chemistry Department has been notified of the impending reactor startup.

2.1.15 The following surveillances have been completed within 7 days I&C Supervisor prior to reactor startup:

SP 2401A, Manual Reactor Trip Test Two of the following (minimum of 2 channels OPERABLE):

SP 2401BB1, Channel A Wide Range Logarithmic Flux Monitor Start-up Functional Test SP 2401BB2, Channel B Wide Range Logarithmic Flux Monitor Start-up Functional Test SP 2401BB3, Channel C Wide Range Logarithmic Flux Monitor Start-up Functional Test SP 2401BB4, Channel DWide Range Logarithmic Flux Monitor Start-up Functional Test SP 2401C, RPS n r b i n e Loss of Load Test SP 2401D, RPS Matrix Logic and Trip Path Relay Test OP 2202 TH1NK Rev. 020-06 4 of 47

3.2 Criticality must be anticipated whenever positive reactivity additions (e.g. CEA withdrawal, dilution, xenon depletion, or temperature changes),

are being made [Ref. 6.31.

3.3 During approach to criticality, operations which could produce a sudden change in RCS boron concentration or temperature, must not be allowed.

3.4 Shift turnover should not occur during approach to criticality. If approach to criticality is in progress, shift turnover must be postponed until one of the following is satisfied [Ref. 6.31:

0 Reactor is critical and power level is stable 0 All CEA regulating rods are fully inserted into the core.

0 The reactor has been shutdown for greater than or equal to 70 hours8.101852e-4 days 0.0194 hours 1.157407e-4 weeks 2.6635e-5 months (xenon free conditions) and criticality is not imminent as indicated by 1/M plots.

3.5 Control Room activities that may distract operators during reactor startup shall be restricted to those necessary for plant operation until reactor startup is complete (e.g., other testing, performance of surveillances, turnover activities, etc.) [Ref. 6.31.

3.6 During reactor startup, conservative actions (Le., boration, CEA insertion, manually tripping reactor to add negative reactivity or reduce power), shall be taken when unexpected situations arise with respect to the following:

0 Reactivity 0 Criticality 0 Any other anomalous behavior of the reactor core 3.7 For any CEA movement, Operators must observe or perform the following:

0 Ensure actual CEA motion is in the direction demanded.

0 Monitor nuclear instrumentation to ensure reactor responds as expected.

0 CHANNEL CHECK indications continuously to ensure decisions are not made based upon a single indication.

3.8 The reactor must be tripped for either of the following conditions:

0 A sustained SUR of 1.0 dpm is achieved.

0 An uncontrolled cooldown occurs Tc less than 500" F [OE Ref. 6.61 Additional actions for uncontrolled cooldown are specified in Attachment 5 3.9 PPC points CVRlA, CVRlB, CVRlC, and CVRlD, NIS LOG PWR LVL, should not be used to determine count rate doubling. The response of these computer points may not indicate a count rate doubling in a timely manner.

Level of Use OP 2202 Continuous STOP Rev. 020-06 7 of 47

4. INSTRUCTIONS 4.1 Establishing Initial Conditions for Reactor Startup L 4.1.1 VERIEY Attachment 8 has been completed and IPTE authorized MTL or TE for release.

L 4.1.2 Refer To Attachment 9 and CONDUCT pre-evolution brief such MTL or TE that all personnel participating in, or affected by evolution, have been informed, at a minimum, of the following:

Objective and expected outcome Initial conditions All precautions All anticipated plant responses Guidance for terminating evolution Any anticipated risks to personnel or equipment Termination and restart criteria Contingency actions 4.1.3 REQUEST Chemistry Department sample RCS and determine present RCS boron concentration.

4.1.4 RESET CEA group upper core stops as follows (PPC master host console terminal):

a. ENTER @JCL:CEA and PRESS RETURN.

b. VERIFY the following is displayed on console typer printout:

Starting G2 CEA (CEA pulse input & dig outputs) *

%RUN -S- PROC-ID, identification of created process is XXXXXXXX (xxxxxxxx = Process identification value generated by PPC) 4.1.5 Refer To Attachment 1 and VERIFY all switches and meters are aligned as specified and indicate properly.

OP 2202 Continuous REVIEW Rev. 020-06 8 of 47

g. VERIFY the following: (front of CPP sections, East DC Switchgear Room)

All CPP red lights lit in accordance with Attachment 6 All ACTM Power green lights and UG Engaged red lights lit CPP Cabinet 1 Section A, ACTM backup power supply breaker is closed.

h. IF any CPP red light is not lit, VERIFY applicable CEDM CPP 240 volt CB breaker in ON (front of CPP sections, inside door).

4.1.8 VERIFY SP 2619A-001, Control Room Daily Surveillance has been performed within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of entering Mode 2 and all acceptance criteria is satisfied, or determined by the SM/US not to limit a Mode change per Technical Specifications.

4.1.9 Refer To OP 2208, Reactivity Calculations, and PERFORM two (2) independent ECP calculations to determine critical boron concentration (RE will perform one AND a licensed Operator will perform the other).

4.1.10 IF ECPs calculation results do not agree within 10 ppm, REQUEST Reactor Engineering perform an evaluation to determine cause of discrepancy and correct ECP calculations.

4.1.11 Refer To OP 2208, Reactivity Calculations and DETERMINE boron concentration required to satisfy SHUTDOWN MARGIN.

4.1.12 RECORD boron concentrations and date-time in Table - 1 .

Required RCS Boron Concentration from Actual RCS Boron Date and Time critical boron (highest) op2208 to Concentration SHUTDOWN MARGIN PPm PPm PPm OP 2202 Continuous Rev. 020-06 10 of 47

4.1.13 VERIFY the following:

Calculated critical boron concentration is greater than boron concentration required to satisfy SHUTDOWN MARGIN.

Actual RCS boron concentration is within 10 pprn of highest calculated ECP critical boron concentration.

h CAUTION

1. Following a boration or dilution, sprays must be forced for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to continuing the reactor startup.

2. Criticality must be anticipated whenever positive reactivity additions (Le. CEA movement, dilution, xenon depletion, or temperature changes), occur or are being made.

~~

4.1.14 ENERGIZE AUDIBLE COUNTRATE monitor and ALIGN to OPERABLE wide range channel (RC-OSE) as follows:

PLACE the CHANNEL SELECTOR7switch to position bL YY LL A , B, C or D for an operable wide range monitor.

(RC-OSE).

VERIFY the count rate signal is audible.

SELECT the desired position on the RATE SCALER, switch (front of RC-OSE).

4.1.15 E sampled RCS boron concentration is not within plus or minus 10 ppm of highest calculated ECP critical boron concentration, PERFORM one of the following [Ref. 6.41:

Refer To OP 2304C, Makeup (Boration and Dilution) Portion of CVCS and ADJUST RCS boron concentration to within 10 ppm of the highest calculated ECP critical boron concentration.

PERFORM the following:

1) NOTIFY Reactor Engineering.

2) Refer To OP 2208, Reactivity Calculations and CALCULATE new ECP using slightly different desired critical CEA position.

4.1.16 VERIFY pressurizer sprays have been forced for greater than or equal to one hour after any boration or dilution, (except for blended makeup) prior to any CEA movement to criticality.

OP 2202 Continuous TH R Rev. 020-06 11 of 47

Attachment 5 Reactor Startup Conditional Actions (Sheet 1 of 3)

1. IF at any time, the following conditions occur, PERFORM the specified action:

I F Tavglowers to between 515 and 525 O F AND the reactor is critical, Refer to OPS Form 2619A-001, Control Room Daily Surveillance and RECORD RCS temperature once every hour.

I F Tav lowers to less than 515 O F AND the reactor is critical, PERFORM the fohowing:

0 RAISE Tavg to greater than 515 OF within 15 minutes.

0 1_F Tavgis not greater than 515 O F within 15 minutes, PLACE plant in HOT STANDBY condition within the next 15 minutes.

0 Refer To T/S LCO 3.1.1.5 and DETERMINE applicability.

I F an uncontrolled cooldown occurs (Tc less than 500 OF), PERFORM the following:

0 reactor and INITIATE EOP 2525, Standard Post Trip Actions.

0 STOP one of the 4 operating RCPs (C-04).

0 Refer To AOP 2558, Emergency Boration, and INITIATE emergency boration.

0 Refer To T/S LCO 3.4.9.1 and DETERMINE applicability.

2. IF at any time a sustained SUR of 1.0 dpm is achieved, TRIP reactor and Go To EOP 2525, Standard Post Trip Actions.

3. IF at any time during reactor startup, it appears that criticality is reached, or is predicted to be reached, outside plus or minus 0.5% A e (0.9% A e for initial startup after refueling) band of ECP, PERFORM the following:

3.1 INSERT all CEA reguluting groups in sequence (C-04).

3.2 REQUEST Chemistry Department sample and determine RCS boron concentration.

3.3 INITIATE a CR for Reactivity Management tracking.

3.4 Refer To OP 2208, Reactivity Calculations and, independent of CEA position, ENSURE adequate SHUTDOWN MARGIN using OPS Form 2208-013, Shutdown Margin Determination.

3.5 NOTIFY Reactor Engineering.

OP 2202 Rev. 020-06 Continuous 37 of 47

Attachment 5 Reactor Startup Conditional Actions (Sheet 2 of 3 )

3.6 Refer To OP 2208, Reactivity Calculations and CALCULATE new ECR 3.7 IF new ECP differs significantly from original ECP (greater than 50 ppm or 0.5% Ae) AND cause has been identified and corrected, RESUME startup based on new ECF!

3.8 E new ECP does not differ significantly (less than 50 ppm or 0.5% Ae) from original ECP, PERFORM the following:

3.8.1 INSERT both CEA shutdown groups.

3.8.2 Refer To OP 2304C, Makeup (Boration and Dilution) Portion of CVCS and INITIATE boration of RCS to HOT SHUTDOWN boron concentration.

3.8.3 Refer To OP 2208, Reactivity Calculations and ENSURE adequate SHUTDOWN MARGIN using OPS Form 2208-013, Shutdown Margin Determination.

3.8.4 NOTIFY SORC to perform an investigation.

3.8.5 Upon SORC review and approval, COMMENCE reactor startup based on Reactor Engineers recommendations.

4. E any of the following conditions exist and RCS boron concentration is less than refueling boron concentration, PERFORM step 4.1 (TS LCO 3.1.3.7):

Less than 4 RCPs are operating RCS temperature is less than 500 O F Pressurizer is less than 2,000 psia High Power trip is not OPERABLE (as specified in Technical Specifications LCO, 3.3.1.1) 4.1 DEENERGIZE CEDMs by one of the following:

OPEN both CEDM MG set output breakers and red TAG to SM 0 DEENERGIZE all coil power programmers and red TAG to I&C Department Manager OP 2202 K Rev. 020-06 Continuous 38 of 47

Attachment 9 Reactor Startup IPTE Items (Sheet 2 of 3)

Engineering Manager Select Engineering Department personnel to perform independent review of IPTE procedure Determines if normal review has fulfilled technical review of IPTE 4 Determines technical adequacy of IPTE to address:

Potential transient Expected plant responses Potential equipment damage Termination criteria On - duty Shift Manager 4 Has authority to release and terminate IPTE Maintains control and responsibility for safe operation of plant ( I R E procedure does not supersede this control or responsibility for safe operation)

Authorizes IPTE for restart following termination Other individuals involved in evolution:

As described in Sections 2.2 and 2.3.

2. PRE-EVOLUTION BRIEF Administrative Requirements:

All personnel involved in evolution, including SM and operating shift personnel, must attend Conducted by MTL and TE (repeat briefings may be conducted by assistant MTL)

Must be presented prior to commencing startup Outline of Brief The MTL shall communicate management expectations, discussing lessons learned from industry experiences, defining responsibilities and authorities, and any results of the validation process The TE shall discuss technical aspects of evolution to include:

All precautions, prerequisites, and initial conditions Reviewing evolution by reviewing procedure and associated documents Potential risks Termination criteria, contingency actions, and restart criteria Expected changes in plant status and deviations from normal plant parameters, setpoints, and limits Summary of restoration or transitions, as required OP 2202 THtNK Rev. 020-06 Continuous 46 of 47

Attachment 9 Reactor Startup IPTE Items (Sheet 3 of 3)

3. TERMINATION CRITERIA IF any of the following plant conditions or parameters are exceeded, plant conditions must be stabilized prior to continuing:

Sustained SUR of 1.0 dpm is achieved Criticality is reached, or is predicted to be reached, outside plus or minus 0.5% A@(0.9% A@for initial startup after refueling) band of ECP Uncontrolled cooldown Actual or potential equipment damage Significant procedure changes are determined to be required A condition is created which causes actual or potential serious personnel injury MTL,TE, or SM determine an unsafe condition exists

4. CONTINGENCY ACTIONS Actions contained in Attachment 5.

Actions contained in Section 3.

5. RESTART CRITERIA Plant conditions have stabilized An investigation has been conducted to determine why condition to terminate evolution occurred WHEN cause of evolution termination is determined and corrected, the following applicable items have been reviewed and performed as necessary:

Procedure modifications and SORC approval Additional briefings, validation, or training Approval from the following:

Station Director: Date:

Supervisor, Nuclear Shift Operations: Date:

On-duty SM: Date:

Another pre-evolution brief is conducted OP 2202 Rev. 020-06 Continuous 47 of 47

Question w71: Attached

References:

Attached OP 2303B SFP Fuel Handling Operations Comments:

Choice B (the correct answer) is an acceptable answer. It correctly describes the required actions if the SFP Platform Crane Operator fails to stop when the stainless steel hose clamp on fuel handling tool is level with the top of SFP platform crane safety rail.

Choice A is also an acceptable response. The stem of the question does not provide any information regarding a human performance error on the part of the SFP Platform Crane Operator (i.e. failure to STOP when the stainless steel hose clamp on fuel handling tool is level with the top of SFP platform crane safety rail). It is reasonable for the examinee to assume that operations are proceeding as expected and that the next action would be to move the bridgekrolley to position the fuel over its final rack location (a move that is not prevented by interlock)

Other considerations of OP 2303B SFP Fuel Handling Operations:

Level of Use Reference; the procedure shall be readily available to the user, in the area where the work activity is being performed, such that the user can obtain a copy of the document as needed to perform the procedure.

Prerequisite 21.2: All personnel participating in fuel handling have been briefed and are thoroughly familiar with this procedure and individual responsibilities.

Examinees were required to answer without the use of reference material.

Recommendation:

Accept Choice B and Choice A Justification:

Examinees who selected Choice B exhibited knowledge of the new and spent fuel movement procedures and also knowledge of fuel handling equipment interlocks.

Examinees who selected Choice A exhibited knowledge of the new and spent fuel movement procedures by correctly identifying the next procedurally directed step, considering that they were not cued that the SFP Platform Crane Operator had incorrectly performed the previous step.

I have reviewed the Comments and attached References for question #7 1 and concur with the

,P recommendation to accept both Choice A d Choice B as correct answers.

/

d RO fl SRO QuestionID: 5000067 1- 1 , d Memory7 (Check=Yes)

A transfer of a new fuel assembly IS in progress from one location in the spent fuel pool to another using OP-2303B, "SFP Fuel Handling Operations" The operator raises the hoist with the desired assembly grappled until upward motion IS stopped by the upper limit switch interlock What must be done next?

A Release hoist raise switch, use the bridgeltrolley controls to move to destination.

B Stop all hoist and crane movement and notify Reactor Engineering immediately C Lower assembly into initial location and contact Reactor Engineering for resolution D Slowly lower hoist until load cell indicates 250 to 290 pounds, then continue move.

Justification CHOICE (A) - NO WRONG: Procedure directs stopping all fuel movement.

VALID DISTRACTOR: Plausible that it is acceptable to have motion stopped by interlock CHOICE (8)- Y E S Danger of ungrappling and dropping fuel assembly. Must stop and notify immediately CHOICE (C) - NO WRONG: Procedure directs stopping all fuel movement.

VALID DISTRACTOR: Plausible that corrective action would be to lower into rack. This is correct action for fuel handling event.

CHOICE (D) - NO WRONG: Procedure directs stopping all fuel movement VALID DISTRACTOR: An applicant may think that use of the interlock affects the load cell. 250 to 290 pounds is the load identified by the procedure for a suspended assembly.

Reference NRC K/A System/E/A NRC WA Generic System 2.2 Equipment Control Number 2.2.28 Knowledge of new and spent fuel movement procedures.

Importance ROlSRO 2.6 3.5 IOCFR Link (CFR: 43.7 / 45.13)

Page 130 of 176

MILLSTONE POWER STATION SURVEILLANCE PROCEDURE

~ ~~

SFP Fuel Handling Operations OP 2303B Rev. 002-01 I TQ F CP VI Ebb!

Approval Date: 2/14/05 Effective Date: 2/ 15/05

4.2.10 WHEN load cell indicates approximately zero pounds, RAISE hoist to obtain 250 to 290 pounds.

4.2.1 1 CLOSE and LOCK grapple.

u- CAUTION v During hoist operations with fuel attached, the upper limit switch interlock must not be used to stop hoist upward motion. If the hoist is raised until the upper limit switch interlock stops hoist motion, notify Reactor Engineering immediately. Do not attempt to lower the assembly into the upender or storage rack.

I I 4.2.12 Slowly RAISE hoist and MONITOR load cell indication.

4.2.13 To prevent running hoist into upper limit switch, WHEN stainless steel hose clamp on fuel handling tool is level with the top of SFP platform crane safety rail, STOP hoist movement.

4.2.14 E the hoist is raised until the upper limit switch interlock stops hoist motion, PERFORM the following:

a. STOP all hoist and crane movement.

b. NOTIFY Reactor Engineering immediately.

\I C A U T I O N v Fuel shall not be lowered below the top of the SFP rack unless one of the following apply: [Ref. 6.91 The fuel assembly is lowered into an approved storage location such as an SFP storage rack cell, NFE, upender, or other approved location.

The fuel assembly remains at least 12 inches away from the nearest fuel assembly when lowered outside an approved storage location.

c. To clear interlock, LOWER hoist per Reactor Engineering direction.

OP 2303B s1OP [<4/1<!< AC 1 14LVIL\A,' Rev. 002 -0 1 11 of49

4.2.15 MOVE SFP platform crane to position fuel assembly over ,final rack location.

4.2.16 Visually VERIFY proper vertical alignment between fuel assembly and funnel.

4.2.17 Slowly LOWER hoist and MONITOR load cell indication.

4.2.18 WHEN fuel assembly is fully lowered into rack, VERIFY load cell indication is between 250 and 290 pounds.

4.2.19 Visually VERIFY fuel assembly fully lowered into rack.

NOTE Fuel handling tool grapple is closed when it is aligned with the tool handles and is open when it is not aligned with the tool handles.

4.2.20 UNLOCK and OPEN grapple.

4.2.21 WHEN grapple is open, RAISE hoist approximately two feet and MONITOR load cell indication.

NOTE Load indication between 300 and 365 indicates an inadvertently withdrawn CEA (if installed).

4.2.22 IF load cell indicates between 300 and 365 pounds, PERFORM the following:

a. NOTIFY Reactor Engineering.

b. Slowly LOWER hoist until load cell indication is between 250 and 290 pounds.

c. VERIFY grapple is in the FULL OPEN position.

d. WHEN grapple is full open, RAISE hoist approximately two feet and MONITOR load cell indication.

e. E load cell indicates between 300 and 365 pounds, STOP fuel movement and CONTACT Reactor Engineering for resolution.

OP 2303B SI 91 E hlKli F C7 HLVII-VV Rev. 002-01 12 of 49

Question ##72:Attached

References:

Attached OP2209A Refueling Operations MP2 Cycle 15 Core Load (1/M vs. Assemblies Loaded)

Comments:

Choice C (the correct answer) is an acceptable answer for an anticipated count rate multiplication due to the loading of a new fuel assembly in a location adjacent to the CH B Wide Range detector. Since no information is supplied in the stem of the question as to the refueling method (e.g. Full Core Reload) it is reasonable to expect this change in some instances.

Choice B is also an acceptable answer. The stem of the question provides no additional information regarding the method of refueling and status of refueling. Assumptions could be made as to the refueling method (e.g. Fuel Shuffle) and previous data trends of a 1/M plot and/or count rate changes.

Historical data shows that during a fuel shuffle the 1 / M value rarely dips below (.8). Using the provided count rate data from the stem of this question 1/M values (CRinitid/ CRfinal)are as follows; CH A (.95), CH B (.56 almost a doubling), CH C (.84),CH D (.83). If an assumption is made that initial count rates at the start of this fuel shuffle was even lower, then its effect on a 1/M plot would be greater and a doubling may be evident. If the operator believes that an unanticipated count rate multiplication has occurred he/she is compelled by OP 2209A to commence an emergency boration.

Recommendation:

Accept Choice C and Choice B Justification:

Choice B correctly describes the required operator action of OP 2209A and is a conservative response to a situation that required judgement.

4.5.15 IF, at any time, unanticipated count rate multiplication, (i.e., doubling), is indicated, PERFORM the following:

a. SUSPEND refuel operations.

b. Refer To AOP 2558, Emergency Boration and PERFORM applicable actions to initiate boration to RCS.

c. Immediately NOTIFY Reactor Engineering and SM.

d. REQUEST evaluation be completed prior to restart of fuel handling activities.

e. INITIATE CR.

I have reviewed the Comments and attached References for question #72 and concur with the recommendation to accept both Choice B and Choice C as correct answers.

A&> +f, 3 1.3 I /os-Stephen F. Claffey, R e a c w n e e r MP2 Date

Refueling is in progress. A new fuel assembly has just been lowered into core location A-1 1 (core map attached). You are the PPO and have noted the following before and after readings on the wide range logarithmic power channels:

BEFORE AFTER WR CH A 1.9E1 cps 2.OE1 cps WR CH B 1.8E1 cps 3.2E1 cps WR CH C 1.6E1 cps 1.9E1 cps WR CH D 1.OE1 cps 1.2E1 cps Based on these indications, which of the following is required?

A Suspend all core alterations and positive reactivity additions.

B Commence boration per AOP-2558, "Emergency Boration".

C Continue to monitor nuclear instruments, NO immediate action required.

D Withdraw the fuel assembly and contact Reactor Engineering for guidance.

Justlflcatlon CHOICE (A) - NO WRONG: Counts have not doubled. The only appreciable increase in counts is on CH I3 which is immediately adjacent to the location of the new assembly.

VALID DISTRACTOR: Per OP-2209A, if at any time, unanticipated count rate multiplication, &e., doubling), is indicated, then suspend refuel operations.

CHOICE (6) NO -

WRONG: Counts have not doubled. The only appreciable increase in counts is on CH B which is immediately adjacent to the location of the new assembly.

VALID DISTRACTOR: Per OP-2209A, if at any time, unanticipated count rate multiplication, (i.e., doubling), is indicated, then commence boration CHOICE (C) - YES Criteria for action is observation of an unanticipated count rate multiplication, (i.e.. doubling). Counts have not doubled.

CHOICE (D) - NO WRONG: Counts have not doubled. The only appreciable increase in counts is on CH B which is immediately adjacent to the location of the new assembly.

VALID DISTRACTOR: Plausible that requirement is to remove the assembly to lower core reactivity while situation is evaluated.

Reference NRC-2005 Question Hlstory NRC WA System/E/A NRC K/A Generic System Equipment Control Number 2.2.30 "Knowledge of RO duties in the control room during fuel handling such as alarms from fuel handling area, communication with fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation."

Importance ROISRO 3.5 3.3 10CFR Llnk (CFR: 45.12)

Page 78 of 109

MILLSTONE POWER STATION GENERAL OPERATING PROCEDURE Refueling Operations OP 2209A Rev. 025-00 Approval Date:

Effective Date:

4.5.15 at any time, unanticipated count rate multiplication, (ie., doubling), is indicated, PERFORM the following:

a. SUSPEND refuel operations.

b. Refer To AOP 2558, Emergency Boration and PERFORM applicable actions to initiate boration to RCS.

c. Immediately NOTIEY Reactor Engineering and SM.

d. REQUEST evaluation be completed prior to restart of fuel handling activities.

e. INITIATE CR.

4.5.16 WHEN transferringfirst fuel bundle(s) between the refuel pool and SFP, REQUEST Health Physics Department survey radiation levels in adjacent CTMT, Aux Building and Enclosure Building areas.

4.5.17 Prior to initiating or restarting any refueling evolutions, VERIFY permission is obtained from Shift Manager.

4.5.18 E performing refueling operations with Containment purge valves closed AND 2-RW-280, REFUEL TRANSFER TUBE ISOLATION, open to maintain Containment pressure within an acceptable range, PERFORM the following:

I

a. ESTABLISH the following temporary PPC alarm values for Containment pressure low range, P8117:

High limit of +6 inWC Low limit of -2.5 inWC

b. MONITOR Containment pressure (C-01, PPC), refuel pool and SFP levels (local).

c. Refer To T/S LCO, 3.9.4 and DETERMINE applicability.

OP 2209A THINK Rev. 025 - 00 23 of 60

a 0) m a

v a

E 0 0 0

'4 9 0 r 0 0

Millstone Unit 2 LOIT 2005 RO and SRO Exam Analysis JUSTIFICATION FOR ACCEPTING 2604AO AND 2308x11 AS A CORRECT ANSWER TO JPM-A2SRO SRO AWO Acceptance.

Millstone Unit 2 is developing a new set of procedures designed to simplify component maintenance and retesting. These new procedures, called Maintenance Operating Procedures (MOPs), are designated as 2300X. The MOPs include the steps of the 2600 procedure and additional sections for venting, draining and tagging for the component in question. When a MOP is approlred, it becomes the preferred post-maintenance procedure and replaces the previously used 2600.

At the time the SRO candidates were trained and when this JPM was developed, the MOPs did not exist. At the time of the NRC Exam, a MOP was approved for the A HPSI Pp. I believe this is the only MOP that is approved for use at this time.

Due to the present state of transition from the 2600 procedures to the 2300X procedures, Millstone Unit 2 requests both 2604AO and 2308x1 1 be considered correct.

@lh Dominion Z l l l l \ t o l l c l <Nuclear Connecticut, Inc. LElVEfj PDominion ON l\\il j t IIlllll I < o p c I i r n 110 l i t --

\ \ ~ r ~ r t o r (~ lI O ( i i h i

?iJR I 3 m5 Mr. J. M. DAntonio 6 APR 7 9 ,?7 :47 Serial No.05-239 U.S. Nuclear Regulatory Commission MPSLicANDB RO 475 Allendale Road Docket No. 50-336 King of Prussia, PA 19406-1431 License No. DPR-65 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 2 0PERAT0 R EXAMINAT ION AD DIT1ONAL INFORMATION In accordance with NUREG-1021, Examination Standard 501, Paragraph C, Section 1a, Dominion Nuclear Connecticut, Inc., hereby transmits as Attachment 1, additional information to support the acceptable performance of JPM A1 RO, Perform an ECP.

If you have any questions or require additional information, please contact Mr. David W.

Dodson at (860) 447-1791, extension 2346.

Very truly yours, w

J AI

&@ice Price President - Millstone

Serial No.05-239 Operator Examination Additional Information Page 2 of 2 Attachments: 1 Commitments made in this letter: None.

cc: Mr. R. J. Conte Chief, Operational Safety Branch U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415 U.S.Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-141 5 Mr. V. Nerses Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North 1 1555 Rockville Pike Mail Stop 8C2 Rockville, MD 20852-2738 Mr. S. M. Schneider NRC Senior Resident Inspector Millstone Power Station U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555

Serial No.05-239 Docket No. 50-336 Attachment 1 JPM A1 RO, Perform an ECP Millstone Power Station Unit 2 Dominion Nuclear Connecticut, Inc. (DNC)

JPM-AIR0 Perform an ECP: Attached

References:

Attached OP 2208-01 1 Rev. 042 Page 1 of 2 Samarium Worth versus Time After Reactor Shutdown from 100% Power Cjxle 16 OP 2208-01 1 Rev. 042 Page 2 of 2 Tabular Data of Samarium Worth versus Time After Reactor Shutdown from 100% Power Cycle 16 OP 2208-019 Rev. 001 Page 1 of 2 Plutonium Buildup Worth vs. Time After Reactor Shutdown Form 100% Power OP 2208-019 Rev. 001 Page 2 of 2 Tabular Data for Plutonium Buildup Worth vs.

Time After Reactor Shutdown Form 100% Power Revised Answer Key for JPM-A 1RO Perform an ECP Page 6 of JPM-AIR0 with pen and ink change.

Comments :

1, The answer key is in error for Estimated samarium worth at criticality and Pu Buildup Worth.

Estimated samarium worth at criticality should be changed to 0.996 => 1.01.

Resulting in Samarium Defect equaling -0.096 => -0.1 1.

Pu Buildup Worth should be changed to ,038 => .046.

Both changes are indicated on the attached answer key

2. Step 6 standards indicates the present boron concentration is 4600 ppm. It should read 460 ppm.

Recommendation:

1. Change the answer key to reflect the above parameters.

2. Change step 6 of the JPM to read 460 ppm.

1

.Justification :

1. The answer key was in error and did not provide an accurate range for those tlvo parameters.

3 Typographical error.

2

JOB PERFORMANCE MEASURE APPROVAL SHEET I JPM

Title:

Perform an ECP ID Number: JPM-A1RO Revision: - 0 II. Initiated:

Richard J. Ashey 1/16/2005 Developer Date III. Reviewed:

-&$jgTL al Reviewer IV. Approved:

plc2 User Department Supervisor Date f l Nuclear raining Supervisor Date

JOBPERFORMANCEMEASUREWORKSHEET Facility: MP-2 Examinee:

JPM Number: JPM-A1RO Rev. 0 Task

Title:

Perform an ECP System: Administrative Time Critical Task: Yes No X Validated Time (minutes): 35 Task No.(s): NUTIMS #121-09-195 Applicable To: SRO X RO X PEO KIA No.: 2.1.25 K/A Rating: 2.813.1 Method of Testinq:

Simulated Performance: Actual Performance: X Location:

Classroom: X Simulator: X In-Plant: X Task Standards: The examinee properly completes an ECP.

Reauired Materials OP 2208, Reactivity Calculations, Rev. 13-05 (procedures,equipment): Completed ECP Data and Analysis Sheet, OP 2208-001 ECP Reference Data Sheet, Attachment 1 General

References:

OP 2208, Reactivity Calculations, Rev. 13-05

* *

READ TO THE EXAMINEE * * *

l will explain the initial conditions, which step(s) to simulate or discuss, and provide initiating cues.

When you complete the task successfully, the objective for this job performance measure will be satisfied. You may use any approved reference materials normally available in the Control Room, including logs. Make all written reports, oral reports, alarm acknowledgments, and log entries as if the evolution was actually being performed.

2 of 18

JOB PERFORMANCE MEASURE WORKSHEET JPM Number: JPM-A1RO Rev. 0 Initiating Cues: The US has directed you to complete an ECP in accordance with OP 2208, Reactivity Calculations, without the use of the PPC.

Initial Conditions: 0 A Reactor startup is planned to begin within the next 30 minutes with criticality anticipated 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months from now.

0 The PPC is presently unavailable, but will be returned to service prior to the startup.

0 Desired critical position is group 7 at 55 steps 0 The following conditions exist:

- The plant tripped from 100% power 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months ago.

- The plant had been at 100% for the past 220 days.

- Present RCS Boron is 460

- Present Burnup provided by Reactor Engineering is 12,988 MWDlMTU

- RCS Tavg is being maintained constant at 532°F.

Reference data taken at 0900 on 03/15/05 at 100% power IS as follows:

- 572Tavg

- 2.794 %Lp Xenon

- 0.900 o/oLp Samarium

- 12,850 MWD/MTU Burnup

- 340 ppm RCS Boron

- Group 7 CEAs at 180 steps Data is good until 14,000 MWD/MTU Simulator Requirements: N/A

* *

NOTES TO EXAMINER * * * *

1. Critical steps for this JPM are indicated with an X.For the examinee to achieve a satisfactory grade, critical steps must be completed correctly.

2. When examinee states what hidher simulated action/observation would be, read the appro priat e Cue .

3. If necessary, question examinee for details of simulated actions I observations (Le. What are you looking at? or What are you observing?).

4. Under NO circumstances must the examinee be allowed to manipulate any devices during the performance of this JPM (in-plant only).

3 o f 18

PERFORMANCE INFORMATION JPM ID NUMBER: JPM-A1RO TITLE: Perform an ECP START TIME:

STEP 1 __ Performance Steps: Verify the following:

Reactor Engineering has completed and provided reference critical position data on Attachment 1, "ECP Reference Data Sheet" (ECP Data Book) 0 Chemistry Department has been requested to sample and.determine present RCS boron concentration.

GRADE- - Standards: Examinee obtains and/or asks for Attachment I, ECP Reference Data Sheet and the RCS Boron concentration.

Cue: 0 Provide a copy of OP 2208-001, ECP Reference Data Sheet and Attachment 1, ECP Reference Data Sheet.

0 If requested, as chemistry, report that the RCS Boron concentration was reported 30 minutes ago.

Comments:

STEP 2 -& Performance Steps: Refer To Attachment 1 and TRANSFER Reference Critical Data.

GRADE X Standards: Examinee records the critical data on O f 2208-001, ECP Data and Analysis Sheet..

Cue:

Comments:

4 of 18

PERFORMANCE INFORMATION JPM ID NUMBER: JPM-AIR0 TITLE: Perform an ECP STEP 3 -X Performance Steps: RECORD the following Estimated Status at Criticality data:

0 Date and time RCS temperature (TAVG)

GRADE- -X Standards: Examinee records the Estimated Status at Criticality is as follows:

0 The dafe is today The time is two hours from now T A V G is 532°F.

Cue:

Comments:

STEP 4 -X Performance Steps: OBTAIN present burnup from one of t h e following and RECORD:

0 CVBURNUP (PPC)

Reactor Engineering.

GRADE X Standards: Examinee obtains Burnup from the initial data sheet and records 12,988 MWD/MTU on the ECP Data and Analysis Sheet.

Cue:

Comments:

5 o f 18

PERFORMANCE INFORMATION JPM ID NUMBER, JPM-AIR0 TITLE: Perform an ECP STEP 5 -X PeFformance Steps: Unless otherwise specified by Reactor Engineering, CHECK core burnup change from reference data specified on Attachment 1, to present burnup, does not exceed 1,000 MWD/MTU.

GRADE X Standards: Examinee determines the change in burnup is less than 1000 M WD/M TU.

Cue: If asked, as Reactor Engineering, there are no additional requirements for the core burnup change.

Comments:

STEP 6 A Performance Steps: WHEN sample results are obtained, RECORD present boron concentration.

GRADE X Standards: Examinee records the recorded present boron concentration on OP 2208-001, ECP Data and Analysis Sheet is 4600 ppm.

Cue: If asked, as chemistry, report that the RCS Boron concentration was reported 30 minutes ago.

Comments:

6of18

PERFORMANCE INFORMATION JPM ID NUMBER: JPM-AIR0 TITLE: Perform an ECP STEP 7 Performance Steps: Refer to Attachment 1 and RECORD Desired Critical CEA Position.

GRADE- X

__ Standards: Examinee records the Desired Critical CEA Position on OP 2208-001, ECP Data and Analysis Sheet. (group 7 at 55 steps).

Cue:

Comments:

STEP 8 2 Performance Steps: DETERMINE Power Defect as follows:

a. RECORD Reference Critical Data power value.

b. Refer to OP 2208-018 and DETERMINE Power Defect at Reference Critical power value.

c. Record Power Defect GRADE X Standards: Examinee records the Reference Critical Data power value on the ECP Data and Analysis Sheet Examinee refers to OP 2208-018 and determines the Power Defect at Reference Critical Power and records it on the ECP Data and Analysis Sheet.

Cue:

Comments:

7 of 18

PERFORMANCE INFORMATION JPM ID NUMBER: JPM-AIR0 TITLE: Perform an ECP STEP 9 X Performance Steps: DETERMINE Xenon Defect as follows:

a. RECORD Reference Critical Data Xenon worth.

b. Refer To one of the following and DETERMINE estimated Xenon worth at criticality:

Xenon-Samarium Post Trip Report (printed automatically on Control Room special typer following reactor trips)

OP2208-004 XENON-SAMARIUM DEMAND program on PPC Reactor Engineering

c. RECORD estimated Xenon worth at criticality

d. CALCULATE Xenon Defect as follows and RECORD:

Reference Critical Data Xenon worth - Estimated Xenon worth = Xenon Defect GRADE- -X Standards: 0 Examinee records the Reference Critical Data for Xenon worth on the ECP Data and Analysis Sheet.

Examinee refers to OP 2208-004 to obtain the estimated Xenon worth at criticality and records the value on the ECP Data and Analysis Sheet.

Examinee records the calculated Xenon Defect.

Cue: The Xenon-Samarium Post Trip Report is NOT available.

Comments:

8 o f 18

PERFORMANCE INFORMATION JPM I O NUMBER: JPM-AlRO TITLE: Perform an ECP STEP 10 -X Performance Steps: DETERMINE Samarium Defect as follows:

a. RECORD Reference Critical Data Samarium worth.

b. Refer To one of the following and DETERMINE estimated Samarium worth at criticality:

Xenon-Samarium Post Trip Report (printed automatically on Control Room special typer following reactor trips)

OP2208-011 0 XENON-SAMARIUM DEMAND program on PPC Reactor Engineering

c. RECORD estimated Samarium worth at criticality.

d. CALCULATE Samarium Defect as follows and RECORD:

Reference Critical Data Samarium worth - Estimated Samarium worth = Samarium Defect GRADE- X

__ Standards: 0 Examinee records the Reference Critical Data for Samarium worth on the ECP Data and Analysis Sheet.

0 Examinee refers to OP 2208-01 1 to obtain the estimated Samarium worth at criticality and records the value on the ECP Data and Analysis Sheet.

0 Examinee calculates the Samarium Defect and records it on the ECP Data and Analysis Sheet.

Cue: The Xenon-Samarium Post Trip Report is NOT available.

Comments:

9 o f 18

PERFORMANCE INFORMATION JPM ID NUMBER: JPM-AIR0 TITLE: Perform an ECP STEP 11 -X Performance Steps: DETERMINE CEA Worth Defect as follows:

a. Refer To OP 2208-007 and DETERMINE CEA Worth for the following and RECORD 0 Reference Critical Data CEA Position 0 Desired Critical CEA Position

b. CALCULATE CEA Worth Defect as follows and RECORD:

Reference Critical Data CEA Position worth - Desired Critical CEA Position worth = CEA Worth Defect G R, . Standards: 0 Examinee refers to OP 2208-007 to obtain the CEA worth for the Reference Critical Data CEA Position and the Desired Critical CEA Posifion.

0 Examinee records the value on the ECP Data and Analysis Sheet.

0 xaminee calculates the CEA Worth Defect and records the value on the ECP Data and Analysis Sheet.

Comments:

I O of 18

PERFORMANCE INFORMATION JPM ID NUMBER: JPM-A1RO TITLE: Perform an ECP STEP 12 X

__ Performance Steps: DETERMINE Boron Defect as follows:

a. RECORD Reference Critical Data boron concentration.

b. RECORD present boron concentration.

c. Refer To OP 2208-005 and DETERMINE the Inverse Boron Worth at present burnup.

d. RECORD Inverse Boron Worth.

e. CALCULATE Boron Defect as follows and RECORD:

(Reference Critical Data boron concentration -

Present boron concentration) / - Inverse Boron Worth

= Boron Defect GRADE X Standards: Examinee records the Reference Critical Data for Boron concentration on the ECP Data and Analysis Sheet.

Examinee records the present Boron concentration of 460 ppm on the ECP Data and Analysis Sheet.

Examinee refers to OP 2208-051 to obtain the Inverse Boron Worth at present burnup and records the value on the ECP Data and Analysis Sheet.

Examinee calculates the Boron Defect and records the value on the ECP Data and Analysis Sheet.

Cue:

Comments:

11 of 18

PERFORMANCE INFORMATION JPM ID NUMBER: JPM-AIR0 TITLE: Perform an ECP STEP 13 -X Performance Steps: DETERMINE Plutonium Buildup as follows:

a. Refer To OP 2208-01 9 and DETERMINE Plutonium Buildup worth at criticality.

b. RECORD Plutonium Buildup worth at criticality GRADE X Standards: Examinee refers to OP 2208-019 to obtain the Plutonium worth at criticality.

Examinee records the value on the ECP Data.

Cue:

Comments:

STEP 14 -X Performance Steps: DETERMINE the sum of all defects as follows:

a. ENTER a// previously calculated reactivity defects.

b. CALCULATE the sum of all reactivity defects and RECORD.

GRADE X Standards: Examinee records each of the defects on the ECP Data and Analysis Sheet.

0 Examinee adds the defects and determines records the total on the ECP Data and Analysis Sheet. (7 6 to -

7 9)

Cue:

Comments:

12 of 18

PERFORMANCE INFORMATION JPM ID NUMBER: JPM-AIR0 TITLE: Perform an ECP STEP 15 -X Performance Steps: DETERMINE Boron Equivalent of Defects as follows:

a. Record the following:

Sum of defects Inverse Boron Worth at present burnup

b. Calculate Boron Equivalent of Defects as follows and record:

Sum of Defects x Inverse Boron Worth = Boron Equivalent of Defects GRADE X Standards: Examinee records the sum of defects and the inverse Boron worth on the ECP Data and Analysis Sheet.

Examinee calculates the value of Boron Equivalent of Defects and records it on the ECP Data and Analysis Sheet.

Cue:

Comments:

STEP 16 -X Performance Steps: DETERMINE Boron Equivalent of Reactivity Change Due to Burnup as follows:

a. If the difference between present burnup and Reference Critical Data burnup is less than or equal to 200 MWD/MTU, record N/A in this section and go to step 4.1.17.

GRADE X Standards: Examinee determines that this step is Not Applicable and enters "N/A" on the ECP Data and Analysis Sheet.

Cue:

Comments:

13 of 18

PERFORMANCE INFORMATION JPM ID NUMBER: JPM-A1RO TITLE: Perform an ECP STEP 17 X

__ Performance Steps: DETERMINE Required Boron Change for Criticality as follows:

a. Record the following:

0 Boron Equivalent of Defects If any, Boron Equivalent of Reactivity Change Due to Burnup.

b. Calculate critical boron concentration as follows and record:

Present Equivalent of Defects + Boron Equivalent of Reactivity Change Due to Burnup = Required Boron Change.

GRADE X Standards: 0 Examinee records the value for Boron Equivalent of Defects on the ECP Data And Analysis Sheet. (-16 to

-19 should be entered.)

0 Examinee determines that the Boron Equivalent of Reactivity Change due to Burnup is N/A.

0 Examinee calculates the Required Boron Change and records it on the ECP Data and Analysis Sheet. (-16 to -19 should be entered.)

Cue:

Comments:

14 of 18

PERFORMANCE INFORMATION JPM ID NUMBER: JPM-AIR0 TITLE: Perform an ECP STEP 18 X Performance Steps: DETERMINE Critical Boron Concentration as follows:

a. Record the following:

Present Boron Concentration Required Boron Change.

b. Calculate the Critical Boron Concentration.:

Present Boron Concentration + Required Boron Change = Critical Boron Concentration.

GRADE X Standards: Examinee records the value for Present Boron Concentration on the ECP Data And Analysis Sheet.

(460 should be entered.)

Examinee records the value for Required Boron Change on the ECP Data And Analysis Sheet. (-16 to

- 19 should be entered.)

. Examinee calculates the critical boron concentration and records it on the ECP Data and Analysis Sheet.

(441 to 444 should be entered)

Cue:

Comments: Per OP-2202step 2.3.1 calculated boron concentration can be + or - 10 ppm.

431 to 454 ppm is acceptable.

15 of 18

PERFORMANCE INFORMATION JPM ID NUMBER: JPM-AIR0 TITLE: Perform an ECP STEP 19 -X Performance Steps: DETERMINE Limits on CEA Position at Criticality as follows:

a. Record the following:

CEA worth for Desired Critical CEA Position.

Calculate CEA worth at minimum insertion.

Calculate CEA worth at maximum insertion.

b. Refer to OP 2208-007 and calculate CEA position at minimum and maximum insertion and record on ECP Data and Analysis Sheet.

Desired Critical CEA Position worth + and - 0.5 %delta rho.

GRADE X Standards: Examinee records the value for Desired Critical CEA Position on the ECP Data And Analysis Sheet. (. 710 to .75 should be entered.)

Examinee records the value for minimum CEA worth on the ECP Data And Analysis Sheet. (.20 to .25 should be entered.)

Examinee records the value for maximum CEA worth on the ECP Data And Analysis Sheet. (1.0 to 1.251 should be entered.)

0 Examinee calculates the CEA position at minimum and maximum insertion and records it on the ECP Data and Analysis Sheet.(Mimimum = Group 7 @ 145 to 155, Maximum = Group 6 @ 55 to 95 should be entered)

Cue:

Comments: The following band is acceptable:

Minimum 155 to 145 steps in Group 7 Maximum 97 to 55 steps in Group 6 Comments: When the examinee completes the ECP, this JPM is complete.

STOP TIME:

16 of 18

VERIFICATION OF JPM COMPLETION Job Performance Measure No. JPM-A1RO Rev. 0 Date Performed:

Operator:

Evaluator(s):

For examinee to achieve a satisfactory grade, ALL critical steps must be completed correctly. If task is Time Critical, it MUST be completed within the specified time to achieve a satisfactory grade.

Time Critical Task? Yes No X Validated Time (minutes): 35 Actual Time to Complete (minutes):

Result of JPM: (Denote by an S for satisfactory or a u for unsatisfactory)

Areas for Improvement:

17 of 18

EXAMINEE HANDOUT JPM ID Number: JPM-AIR0 Initiating Cues: The US has directed you to complete an ECP in accordance with OP 2208, Reactivity Calculations, without the use of the PPC.

Initial Conditions: A Reactor startup is planned to begin within the next 30 minutes with criticality anticipated 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months from now.

The PPC is presently unavailable, but will be returned to service prior to the startup.

Desired critical position is group 7 at 55 steps The following conditions exist:

- The plant tripped from 100% power 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months ago.

- The plant had been at 100°/~for the past 220 days.

- Present RCS Boron is 460

- Present Burnup provided by Reactor Engineering is 12,988 MWD/MTU

- RCS Tavg is being maintained constant at 532°F.

Reference data taken at E900 on 03115/05 at 100% power is as follows:

- 572"Tavg

- 2.794 %Ap Xenon

- 0.900 % I A Samarium

~

- 12,850 MWD/MTU Burnup

- 340 ppm RCS Boron

- Group 7 CEAs at 180 steps 0 Data is good until 14,000 MWD/MTU 18 of 18

7-NN r - t . 4 -

0 00 0 ml i

c "

d I

- I I

I I

I 1

4

-I I

I L

1 :

Tabular Data of Samarium Worth vs. Time After Reactor Shutdown From 100% Power Cycle 16 Time After 150 1000 5000 9000 13000 15705 Shutdown MWD/MTU MWD/MTU MWD/MTU MWD/MTU MWD/MTU MWD/MTU (hours) 0 0.541 %Ap 0.775 %Ap 0.830 %Ap 0.867 YoAp 0.900 YoAp 0.920 %Ap 15 0.589 %Ap 0.835 %Ap 0.894 '/oAp 0.936 YoAp 0.975 %Ap 1.OOO YoAp 25 0.61 5 %Ap 0.870 YoAp 0.929 %Ap 0.975 %Ap 1.019 %Ap 1.045 %Ap 50 0.669 %Ap 0.937 %Ap 1.002 YoAp 1.052 YoAp 1.I03 ToAp 1.137 %Ap 75 0.708 %Ap 0.987 %Ap 1.054 %Ap I.IO9 YoAp 1.I65 %A.p 1.202 %Ap 100 0.735 %Ap 1.O22 %Ap 1.091 YoAp 1.I49 %Ap 1.210 %Ap 1.249 %Ap 150 0.769 YoAp 1.066 YoAp 1.I36 %Ap 1.I99 %Ap 1.265 %Ap .307 YoAp 200 0.788 %Ap 1.088 %Ap 1.161 %Ap 1.225 %Ap I.293YoAp 1.338 %Ap 300 0.802 %Ap 1.105 %Ap 1.181 %Ap 1.246 %Ap I.315 YoAp 1.362 %Ap 500 0.806 %AD 1.111 YoAp 1.I86 YoAp 1.253 %Ap 1.323 %Ap 1.370 %Ap OP 2208-011 Rev. 042 Page 2 of 2

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Tabular Data for Pu-239 Buildup Worth versus Time After Reactor Shutdown From 100% Power Cycle 16 I ;zd2; I 0 hour0 days 0 hours 0 weeks 0 months s-II 150MWD'MTU 0.000 %Ap 9000 MWD/MTU 0.000 YoAp 15705 MWD/MTU 0.000 YoAp 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months 0.024 %Ap 0.032O/oAp 0.033%Ap 25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months 0.038 %Ap 0.050 YoAp 0.052 YoAp 50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months 0.067 %Ap 0.086 %Ap 0.089 %Ap 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months 0.1 52 O/oAp 0.181 %Ap 0.1 86 YoAp 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months 0.171 %Ap 0.205 YoAp 0.21 O '/oAp 200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months 0.180 YoAp 0.216 OhAp 0.222 %Ap 300 hours0.00347 days 0.0833 hours 4.960317e-4 weeks 1.1415e-4 months 0.187 %Ap 0.224 %Ap 0.229 YoAp 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months 0.1 86 YoAp 0.222 YoAp 0.224 YoAp OP 2208-019 Rev. 001 Page 2 of 2

Form Approval 1 Approval Date Effective Date 10/18/04 12/ 16/04 I Controlling Regulating Group 7 at &steps Estimated Status at Criticality Datemime Burnup (present)

T~~~

N O U + 2/43. 532 "F I 2 , gag MWD/MTU.

'Desired Critical CEA Position Boron (present)

CEAGroup Jat s- steps /60 DDm Reference Data power Power Defect at Reference Critical Data Power Defect

/OD %

power value (OP 2208-018) x",O/ 5 2.33 %Ae Xenon Defect I

ISamarium Defect

~

1 OP 2208 -00 L Rev. 007-01 Page 1 of 3

- -~ ~ . _- __ -- __

Plutonium Buildup Worth Plutonium Buildup Worth at Criticality (OP 2208-019) Pu Buildup Worth !

I\

"I I %Ae IBoron Equivalent of Defects Boron Equivalent of Reactivity Change Due to Burnup (SP 2101s-002)

(N/A if present burnup minus Reference Critical Data burnup is 5 200 MWD/MTU)

I oncentration I Required Boron Change For Criticality I Critical Boron Concentration I L713 O P 224 I Limits on CEA Position at Criticality ~ -1 Desired Critical CEA Position worth I 0

OP 2208 -00 1 Rev. 007 -0 1 Page 2 of 3

ECP C;ilcul;itetl B),( s i g a t i i r e ) : Date: Tiiiic:

I 1

Approved By SCI/US!RE: Date. Tim:

1 I

I I Daterrime TAVG Critical Number "F

CEA Position Boron I1 CEA Group __ at steps PPm

~

Remarks:

I j

I 1

I I

I i

I 1

i Reviewed By Reactor Engineer: Date:

I OP 2205-001 Rev. 007 - 0 1 i Page 3 of 3

PERFORMANCE INFORMATION JPM ID NUMBER: JPM-A1RO TITLE: Perform an ECP STEP 5 -X Performance Steps: Unless otherwise specified by Reactor Engineering, CHECK core burnup change from reference data specified on Attachment 1, to present burnup, does not exceed 1,000 MWDIMTU.

GRADE X Standards: Examinee determines the change in burnup is less than 1000 M WD/M TU.

Cue: If asked, as Reactor Engineering, there are no additional requirements for the core burnup change.

Comments:

STEP 6 X Performance Steps: WHEN sample results are obtained, RECORD present boron concentration.

GRADE X Standards: Examinee records the recorded present boron concentration on OP 2208-001, ECP Data and Analysis Sheet is d606ppm.

460 //fir Cue: If asked, as chemistry, report that the RCS Boron concentration was reported 30 minutes ago.

Comments:

6 o f 18

FAX NO. a60 437 mi P. 31 Nti ominion" Fax TransmittaI To:

Company: hlRC Department: I Fax:

From:

rl\iS fur i s hlcnded lor die fcciphl or artily abovc. 11 m y conlain iafomlion tlirrl is privileged. confidentid or work-produn domain. lf the reader of this meusy is no( ihe intended d p i c n t , or ihc employee responsiblefor delivering this communicafion IO Lhc inmdod recipient, you we hereby notified thiu any disclon R. disuibuiion or copyingof h i s mmmunica1jon is strictly prohibird. I f you have rcccived Ibis cornrnwidon in mor.please imrncdiricty nolify us by lclcphonc so we can a m p for its rcium. Thank you.

Comments: I

FPR-21-cbd> T:IU 12:54 PM 30MINI3N TRAINING

^-?I-FAX NO. 860 137 257: ?, ;2 SPM-A2RO: Atrached

References:

None Recoinmenda tion:

Change the JPM to exclude tagging the 13 !hiller Compressor Breaker. 82174, as a critical srcp.

Justifica tioa:

The JPM requires the operator to provide rec ominended tagging in order to perfoim maintenance on a valve in the Chilled Water iystein. The C Chilled Water Pump must be included in this tag out to prevent the flow of ChilledWater to the valve bcing removed for replair. . The Chiller Compressor Breaker, B2 .74,is tagged to conseivntively provide additional equipment protection (NOT persoixd protecti In) because the CChilled Water Pump is normally aligned to supply Chilled Water to 1 le 13 Chiller Compressor. Securing Chilled Warer LO a Non-Vital Chiller Compressor wil automatically stop the compressor. The C Chilled Warn Pump must be tagged in the pr ,per sequence for the staled schediiled maintentmce. Interlocks are provided to prev n t the B Chiller Compressor irom starting if Chilled Water flow or a Chilled Water Pump s NOT available. Even if the interlocks preventing the start of the compressor were to fail, other irotective features in the compressor circuitry would trip the compressor to prevent equipmc nt damage. Tagging the B Chiller Compressor in a specific order would only be required if K laintenance were being performed that would impact the function of the interlocks. Tn tlis =e, the interlocks will remain in effect. The B (swing) Chilled Water m p is NOT availabl I to supply the B Chiller Compressor due to tagging the cross-tie valve, 2-CI-IW-125; therr fore, the Compressor is ragged as a conservative measure to ensure the compressor will NOT 5 art in the event of a failure of its inierlocks. The tagging sequence is irrelevant due the niirnerc i s interlocks and protective features associated with a loss of Chilled Water to the compress0 In addition, Chilled Water Sysiem, OP 2330C does NOT require the Non-Vital Compressors to be tagged when securing and/or draining the :ystein for maintenance; therefore, tagging sequence is irrelevant.

I have reviewed the Recommendation and Jus ification For JPM-AZRO and concur with the recommendation to exclude tagging the BC hiller Compressor Breaker, 132174, as a critical step. J also agree that h e sequence for taggin, ; the Chiller Compressor Breaker, B2174 is NOT

/

hifi Operations MP2 Date

APR-21-2035 THU 12:55 PM DOMINION TRAINING FAX NO, 860 437 287i JOB PERFORMANCE M iASURE APPROVAL SHEET I. JPM

Title:

RO Taq Clearance PFe aration ID Number: J PM-A2RO Revision: -0 II. Initiated:

A. Pantalon 1/28/05 Developer Date III. Reviewed:

JP4- Technic I Reviewe IV. Approved:

rij4 User Department Super sor Date

or Date

APR-21-2305 THU 12:55 PI DOM NION TRA NING FAX NO, 860 437 267: ?, c4

SUMMARY

OF CHANGES I A/l&Date I DE! RIPTION REVlCHANGE I 11-15-2005 (DAW I Developed new JPM 0 Deleted tagsglng Chlller Compreseor Bred 82174, as a critical step. 011 9

2

A?R-2;-2:35 THU 12:55 ?M DOMINI3N TRAINING FAX NO, 860 437 257i P, 05 I

JOB PERFORMANCE MEASURE WORKSHEET Facility: MP-2 Examinee:

JPM Number: JPM-A2RO Rev. 0 System: Administrative Time Critical Task: Yes -No X Validated Time (minutes): 25 Task No.(s): NUTIMS #119-03-170 Applicable To: SRO x RO X PEO X K/A No.: 2.2.13 WA REting: 3.6/3.8 Method of Testina:

Simulated Performance: ctual Performance:

Location:

Classroom: X Task Standards:

Rewired Materials General

References:

3

LP?.-2!-20,5 7F.L i 2 : 53 ?K DOMINION TRAINING FAX NO. 860 437 2671 P. OE JOB PERFORMANCE dEASURE WORKSHEET JPM Number: JPM-A2RO Rev. 0 I

lnitlatina Cues: The WC-SRO has reviewed the work package and has directed The Tagout N Jmber (Section Number) is 233OC62-003 The AWO nurlber is M2-04-03686 The contact p m o n is V. Team Chemistry will advise as to the disposal of water per NPDES.

All drawings have been verified "Controlled, Approved, and Up to Date".

The examiner will review and approve the tagout.

bitia) Conditions: I Water Pump (P-l49C) Discharge Isolation" al and replacement of the valve.

ter Program is unavailable.

has been evaluated and approved the computer as soon as the Simulator Requirements: None

*

MOTES TO EXAMINER * *

1. Critical steps for this JPM are indicated 'Nith an 'Xu. For the examinee to achieve a satisfactory grade, & critical I&steps mLst be completed correctly.

2. When examinee states what hidher simulated actlordobservation would be, read the appropriate 'Cue'.

3. If necessary, question examinee for deti ils of simulated actions / observations (Le. 'What are you looking at?" or "What are you otiserving?").

4. Under NO circumstances must the exariinee be allowed to manlpulate any devices during the performance of this JPM (in-plant orly).

APR-21-2005 THU 12:55 PM DOMINION TRAINING FAX NO, 860 437 2G71 ?, 07 PERFORMANCE INFORMATION JPM ID NUMBER; JPM-A2RO TITLE: RO Tag Clearance Preparatlon STEP 1 -X Performance Steps: Per WC-2, Attachment 8 , Manual Tagouts

- Prepare a Manual Tagout using (Attachment 9),Manual Tagout Sheer, and appropriate sections in this procedure.

- r3P-2330C-001 Chilled Water System Valve Alignment The examinee deiermines the:

- Zomponents being tagged.

- ?olors for each tag.

Comments: The examlnee may use equh alent documentation for determining components to be tagged.

1--1-.-1--____111_______^

APR-21-2035 TIU 12:56 PI DOMINION TRAINING FAX NO. 860 437 267: P. 08 PERFORMANC'EINFORMATION JPM ID NUMBER: JPM-A2RO TITLE: RO Taq Clearance Preparation STEP 2 -X Performance Steps: E iter the sequential steps for establishing a safe w irking area to include:

Components

- Equipment ID ,

Tag color

- Appropriate instructions Appropriate Tag Position GRADE ent 9,enter the following:

Comments: Special lnstructio s involvement in draining fluids and/or NPDES.

6

E

?,PR-21-2G:5 Til: :2:56 PY DOMINION TRAINING FAX NO. 866 437 2E7! P. 09 JPM ID NUMBER: Preparation STEP 3 GRADE X

Standards:

i Performance Steps: E ter the component name, identification number, lo ation, tag type, and required position on A achment9.

wing information on Attachment w, P149C-HS: Non-Vital Chilled Water Pump rbine Bullding 746, (Optional) 82175 P749C Non-Vital Chilled Water Pump C Comments: The component order may slightly provided:

- Breaker 82175 is before the boundary valves are closed, (Le.,

prior to closing the pump dlscharge and the vent and drain valves.

hP?.-21-2OSZ 7EU 12:s; ?E DOMINION TRAININI FAX NO, 860 437 2G71 PER FORMAI E INFORMATION JPM ID NUMBER: JPM-A2RO TIT RO Tas Clearance Preparation STEP 4 - Performance Steps: mplete block number 11, "Prepared by:"

GRADE Standards: Exan 30 enters his (her) name in block 1 7 of Aftac lent 9.

Comments: Afler thls step is complet

?.PR-21-1u,? Til;' 12:56 PI DOMINION TRAINING n n m -

FAX NO, 860 437 2 F ;

VERlFlCATlON C JPM COMPLETION Job Performance Measure No. JPM-A: to Rev. -0 Date Performed:

Operator:

Evaluator(s):

For examinee to achieve a satisfactory grade, 4 : critical steps must be completed correctly. If task is Time Critical, it MUST be completed wlthin the s cified time to achieve a satisfactory grade.

Time Critical Task? Yes No -

Validated Time (minutes): 25 Actual Tlme to Complete (minutes):

Result of JPM: (Denote by an S Ir satisfactory or a u for unsatisfactory)

Areas for ImProvement:

9

P,?R-21-2;:5 Td3 :2:56 PM DOMINI3N TRAINING FAX NO, 860 437 2671 P. 12 JPM Number: Rev. 0 Initiatinci Cues: eviewed the work package and has directed out for 2-CHW-123 Chill Water Pump (P-o leakage past the seat.

ation is not required.

er (Section Number) Is 2330662-003 s M2-04-03686 Initial Conditions:

10

TmG ONLY Manual Tagout Sheet ANSVVERKEY 1 I (Sheet 1 of 1)

?hisforin is for nrnnirol icse only, it is not inrended to nintch o conipirter generated form I. Tagout Number

!?30C62-003

5. Equipmcot 1 2. Date Today I 3 . AWO Number ("or Ivldtiple") 4. Contact Person w - 0 4 - 0 3 686 V . Team

'Tqout Number

8. T q Lift Sheet Attached [Comm. 3.41 __ Yes

) - C W - 1 23 "Chill Water Pump (P-147C)-Discharge Tsolatian" _ _ _ _- __ 9. Additional AC\fOs uudcr h s tagout __ Yes

6. Reasou ' l ~ g e d IO. Purtial Restoration __ Yes e 2 - C W - I V . Tx-eat.

7. Special InstructiondCaution 1 1. Prepared by Zontact chemktry for draining instructions and NPDES considerations. Examinees Name 12a 12c. Tag Placed Independent Able: lrrilial p0siho)ifar Dhe be^ ib7.4 ifpositbi riot required Step Action complete Verification 8 Red: 2c'WiV-147. "Chill Water Pump (P-149C) Discharge Drain Turbine Buildhe 14'6" (Open) I 9 Red: 2-CW-174. "PI 49C Cmsstie Header Vent", Turbine Building 14'6" (Open)

I 13a. Tagout correct and equipment may bc isolated SWUS notified for power block Boundnry approved by:

AuthuriLttd to be hung by:

%e tv SR no4 ~ ~ l 3d sCLp OC +kd s p m .

Information 1 WC 2 Rev. 006-06 76 of 86

Post Exam Comments Millstone Unit 2 March 2005 Initial License Exam An NRC-developed written exam was administered to 7 RO applicants and 2 SROl applicants on Friday March 18, 2005. The facility-developed operational exam was administered to these same applicants during the following week (3/21/05 - 3/24/05).

The facility submitted post exam comments for the following 5 questions on the RO written exam and also for the following job performance measures in the admin area:

Written Exam Question Number Admin Job Performance Measure 32 A2SRO 37 A1 RO 66 A2RO 71 72 The facility-proposed change to Question #32 on the written exam is accepted as proposed. All other facility-proposed changes to written exam questions (#38, 66, 71 and 72) are not accepted. The facility-proposed change to the Admin JPMs are accepted. The final facility comment was received on April 21,2005.

1. Written Exam Question #32 Question:

The plant is operating at full power with all equipment functional, except for the 'B' HPSI Pump, which is 00s for maintenance.

Then, a large break LOCA occurs combined with a loss of Bus 24D (due to an electrical fault on 24D).

Which one of the choices correctly completes the following statement regarding the impact of the loss of ECCS pumps.

10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months after the event, a loss of the only available adversely affect long term core cooling because the remaining A HPSI pump would, LPSI pump does NOT have a system flowpath for boron precipitation control.

B HPSI pump would, LPSI pump could NOT be procedurally realigned for boron precipitation control via hot leg injection. [Key answer]

C LPSI pump would NOT, HF'SI pump is preferred for boron precipitation control.

D LPSI pump would NOT, HPSI pump could be procedurally realigned for boron precipitation control via hot leg injection.

' Facilitv Comments:

Comments:

The justification for the original correct choice 'B' states that although it is physically possible to align 'A' Page 1 of 11

Post Exam Comments Millstone Unit 2 March 2005 Initial License Exam LPSI pump for Hot Leg Injection, there is no procedural guidance to do so. The reason there is no procedural guidance for this system alignment under the given conditions (24D deenergized), is because without 24D power it is not possible to align LPSI for Hot Leg Injection. One of the major valves in the flow path to achieve this alignment (2-SI-652) is powered by Facility Two (24D) and is located in containment (inaccessible with a LOCA). The procedure section the operators are referred to with 24D lost assumes Facility One HPSI & LPSI pumps are both available.

Therefore, as stated in Choice A, if the available HPSI is lost, core cooling would be in jeopardy because there is no system flowath for Hot Leg Injection with the available LPSI pump.

Recommendation:

Based on the above explanation, we believe both Choice A and Choice B are technically correct for the given information in the stem.

NRC Response:

Facility recommendation for multiple correct answers is ACCEPTED.

The question stem states that Bus 24D is de-energized. Motor-operated gate valve, 2-SI-652, is in the flowpath for LPSI hot leg recirculation and must be opened to set up that flowpath. However, the valve operator is powered from the de-energized bus and therefore 2-SI-652 cannot be opened during the given event. These facts establish the conditions necessary for Choice A to be a correct answer for the question.

Choice B remains an equally valid correct answer. Choice A and Choice B are both correct.

References:

- EOP-2541, Appendix 18, Simultaneous Hot and Cold Leg Injection, Revision OOO

- SDC-00-C, Shutdown Cooling System Lesson Material, Revision 3

2. Written Exam Question #38 Question:

Given the following plant conditions:

- 100% power

- SG levels at setpoint

- Steam flow and feed flow matched

- SG2 Feed Flow Transmitter FT-5269A output fails high With NO operator actions, which of the following describes the expected plant response?

A SG level lowers, but stabilizes above the low level reactor trip. [Key answer]

B SG level lowers to the low level reactor trip.

C SG level rises, but stabilizes below the high level turbine trip.

D SG level rises to the high level turbine trip.

Page 2 of 11

Post Exam Comments Millstone Unit 2 March 2005 Initial License Exam Facilitv Comments:

Comments:

Our original concern with this question was that it required the Candidate to go beyond the knowledge solicited by the WA (i.e.; how will the system respond to the failed instrument), and make a quantitative judgment as to the amount the system will respond to the failed instrument. That was the reason behind our suggested rewrite of the question (see attached) to eliminate the choice of ...low level reactor trip and replace this distractor with one that was clearly wrong in its magnitude of response. In disallowing this suggested change, the Candidate was forced into a quantitative judgment, which depending on the specific tuning of the system for that operating cycle, could result in either Choice A or Choice B being the correct response.

Recommendation:

Based on the above explanation, we believe both Choice A and Choice B are acceptable answers for the given information in the stem.

NRC ResDonse:

Facility recommendation for multiple correct answers is NOT ACCEPTED.

Question matches the selected WA in that it tests an applicants knowledge of the design response to a failure high of a feed flow input to the SG level control system, which controls feed regulating valve position.

Selected WA: 059.K4.08 Knowledge of MFW design feature(s) andor interlock(s) which provide for the following: Feedwater regulatory valve operation (on basis of steam flow, feed flow mismatch)

The facility concern would be understandable if whether or not the plant tripped for this malfunction a borderline situation, but it is not - this malfunction does not approach the trip setpoint. It is reasonable and appropriate to expect a license applicant to be able to have some idea of the magnitude of pant response to an instrument failure, and in this case to identify whether or not this particular SGWLC instrument failure will result in an automatic trip. In order to correctly answer the question, the applicant must understand that the outputs of two feed flow transmitters are averaged to become a single input to the steam / feed mismatch portion of the three-element level control system. Further, the applicant must realize that at 100% reactor power each feed flow transmitter output is near the high end of its instrument range (100% power reading 5.9 E6 lbs/hr, with full range reading 6.OE6 lbshr). With this information an applicant can easily determine that a failure high of a feed flow transmitter will cause a relatively small change to the averaged feed flow signal. The feed regulating valve will begin to throttle closed in response to the change. However, the level input signal will cause the feed regulating valve to reopen with only a small change in actual level.

The facility staff ran a failure high malfunction on one feed flow transmitter for #2 SG on the simulator from 100% power for demonstration purposes. No operator action was taken. #2 SG level was observed to drop from 70% NR level to 67.5% NR level over approximately 2 minutes before turning and beginning to trend back to setpoint. No reactor trip occurred. The low level reactor trip is set to occur when SG level lowers to 49.5% NR level (TS requires setpoint >48.5%).

Given that the failure only results in a 3% change in SG level and that a 20.5% to 21.5% level change Page 3 of 11

Post Exam Comments Millstone Unit 2 March 2005 Initial License Exam would be required to initiate the reactor trip, the distractor B answer (that level will lower to the low level reactor trip) is clearly wrong in its magnitude of response and could not be a correct response regardless of specific tuning for that operating cycle.

References:

- MP2 Tech Specs, Section 2.2. Limiting Safety System Settings

- RPS-01-C, Reactor Protection System Lesson Material Revision 6

- FWC-01-C, Feedwater Control System, Lesson Material Revision 2

3. Written Exam Question #66 Question:

Unit 2 is conducting a reactor start up. Given the following events and conditions:

- Wide range (WR) logarithmic nuclear instrument (NI) channels C and D are out of service

- The reactor is not yet critical

- The ECP expected critical rod height is 100 steps on Regulating Group 6

- Regulating Group 4 is withdrawn to 60 steps

- WR NI Channel A failed low WRL NI Channel A <l.OE-l CPS WRL NI Channel B 6.2E2 CPS Which one of the following statements correctly describes the required action (if any) required to comply with TECHNICAL SPECIFICATIONS?

A Immediately trip the reactor.

B Insert all control rods and shutdown the reactor.

C Stop the startup until WRL NI Channel A has been repaired. NO other actions are required.

D Immediately ensure adequate shutdown margin. [Key answer]

Facilitv Comments:

Comments:

Choice D(the correct answer) is directly related to ACTION 4 for verifying compliance with TS 3.1.1.1 SHUTDOWN MARGIN, which is applicable for the current mode of operation (MODE 3). Choice D is an acceptable answer but is not a complete answer, taken by itself, because it implies that startup may continue.

Choice B Insert all control rods and shutdown the reactor is also an acceptable response because it is a proceduralized action of OP 2202 Reactor Startup IPTE and is the conservative philosophy of operations in DNAP- 1410 Reactivity Management.

These actions prevent non-compliance with TS LCO 3.0.4and its BASIS, which requires exercise of good practice in restoring systems or components to OPERABLE status before plant startup.

The Examinee may have been more compelled to place emphasis on compliance with reactor startup Page 4 of 11

Post Exam Comments Millstone Unit 2 March 2005 Initial License Exam termination than shutdown margin verification because the startup procedure had just previously verified adequate SDM and has controls in place to continuously monitor conditions that could affect SDM (e.g.

ECP). It would be unacceptable to maintain current plant conditions considering the time required to re-verify SDM by obtaining and analyzing an RCS boron sample.

Recommendation:

Accept Choice D and Choice B NRC Response:

Facility recommendation for multiple correct answers is NOT ACCEPTED.

The facility is reading implications into the distractors which are not stated, and also appear to be reading implications of required actions ( insert rods to shutdown the reactor) into their procedures when such actions are not stated. This question specifically asks the required actions to satisfy TECHNICAL SPECIFICATIONS. The facility desired second correct answer insert all rods and shutdown the reactor does not satisfy the Technical Specification requirement to verify shutdown margin. Inserting rods is most certainly a prudent action, but that action is not explicitly required by any procedure referenced by the licensee as a response to loss of WRNIs ,does NOT satisfy the TS, and recognition of that fact was the whole point of this question. The requirement for wide range logarithmic nuclear instruments in MODE 3 per Technical Specifications is for a minimum of 2 OPERABLE channels. With less than 2 OPERABLE channels, TS 3.3.1 Action 4 directs immediate verification of shutdown margin requirements. While there are additional actions that may be taken as a prudent response to inadequate instrumentation during a reactor startup, these actions are not required to comply with technical specifications, and are not explicitly required by any facility procedure. Further, Choice B cannot be considered a correct answer to the question because it does not include the actions required by technical specifications.

References:

- MP2 Technical Specifications, Table 3.3-1, Reactor Protective Instrumentation

4. Written Exam Question #71 Question:

A transfer of a new fuel assembly is in progress from one location in the spent fuel pool to another using OP-2303B, SFPFuel Handling Operations. The operator raises the hoist with the desired assembly grappled until upward motion is stopped by the upper limit switch interlock.

What must be done next?

A Release hoist raise switch, use the bridgdtrolley controls to move to destination.

B Stop all hoist and crane movement and notify Reactor Engineering immediately. [Key answer]

C Lower assembly into initial location and contact Reactor Engineering for resolution.

D Slowly lower hoist until load cell indicates 250 to 290 pounds, then continue move.

Facilitv Comments:

Page 5 of 11

Post Exam Comments Millstone Unit 2 March 2005 Initial License Exam Comments:

Choice B (the correct answer) is an acceptable answer. It correctly describes the required actions if the SFP Platform Crane Operator fails to stop when the stainless steel hose clamp on fuel handling tool is level with the top of SFP platform crane safety rail.

Choice A is also an acceptable response. The stem of the question does not provide any information regarding a human Performance error on the part of the SFP Platform Crane Operator (Le. failure to STOP when the stainless steel hose clamp on fuel handling tool is level with the top of SFT platform crane safety rail). It is reasonable for the examinee to assume that operations are proceeding as expected and that the next action would be to move the bridgdtrolley to position the fuel over its final rack location (a move that is not prevented by interlock).

Other considerations of OP 2303B SFP Fuel Handling Operations:

Level of Use Reference; the procedure shall be readily available to the user, in the area where the work activity is being performed, such that the user can obtain a copy of the document as needed to perform the procedure.

Prerequisite 2.1.2: All personnel participating in fuel handling have been briefed and are thoroughly familiar with this procedure and individual responsibilities.

Examinees were required to answer without the use of reference material.

Recommendation:

Accept Choice B and Choice A Justification:

Examinees who selected Choice B exhibited knowledge of the new and spent fuel movement procedures and also knowledge of fuel handling equipment interlocks.

Examinees who selected Choice A exhibited knowledge of the new and spent fuel movement procedures by correctly identifying the next procedurally directed step, considering that they were not cued that the SFP Platform Crane Operator had incorrectly performed the previous step.

NRC ResDonse:

Facility recommendation for multiple correct answers is NOT ACCEPTED.

Question matches the selected K/A in that it tests an applicants knowledge of an important operational restriction contained within spent fuel handling procedures.

Selected KIA: Generic 2.2.28 Knowledge of new and spent fuel movement procedures.

Facility comments that it is reasonable for an applicant to assume that operations are proceeding as expected since the question stem does not state that a human performance error has occurred. However, the applicants ability to identify that fuel movement being stopped by an interlock is an abnormal situation is central to the knowledge being tested. It is not an assumption, but a determination the applicant was expected to make based on the conditions in the stem of the question. An applicant with the expected level Page 6 of 11

Post Exam Comments Millstone Unit 2 March 2005 Initial License Exam of knowledge regarding fuel handling operations and, in particular, knowledge of important precautions contained in OP-2303B, "SFPFuel Handling Operations", would recognize that the described actions are prohibited and that the procedure requires immediate notification of Reactor Engineering.

Further, the statement "do not attempt to lower the assembly into the upender or storage rack" is contained in both the precaution (3.1) and in the caution prior to the step for raising the assembly (4.2.12). Step 4.2.14 directs the operator to stop all hoist and crane movement and notify reactor engineering immediately if the upper limit switch interlock stops hoist motion.

The facility's contention that Choice A is also an acceptable response is incorrect. There is no ambiguity within the procedure regarding required actions when upward motion of a grappled assembly is stopped by the upper limit switch interlock.

The facility also challenges this question because fuel handling operations would be briefed and the applicant would have the procedure readily available. This question does not require an applicant to have the procedure memorized. This question tests understanding of an important operating limitation appropriately emphasized within a procedure. To answer this question, the applicants must recognize that actuation of the interlock was an abnormal situation, and that the response to abnormal occurrences during fuel handling is to notify RE. This is not an unreasonable level of detail to expect from an applicant on a closed reference question. While the examination standard does allow use of reference materials on a selective basis as attachments to the written exam, it cautions that the references must not improve the applicant's chances of guessing the correct answer by eliminating incorrect distractors. Use of a reference in this instance would have reduced this question level of difficulty to that of a direct lookup.

5. RO Written Exam Question #72 Question:

Refueling is in progress. A new fuel assembly has just been lowered into core location A- 11 (core map attached). You are the PPO and have noted the following before and after readings on the wide range logarithmic power channels:

BEFORE AFTER WRCHA 1.9E1 cps 2.OE1 cps WRCHB 1.8E1 cps 3.2E1 cps WRCHC 1.6E1 cps 1.9E1 cps WR CH D 1.OE1 cps 1.2E1 cps Based on these indications, which of the following is required?

A Suspend all core alterations and positive reactivity additions.

B Commence boration per AOP-2558, "Emergency Boration".

C Continue to monitor nuclear instruments, NO immediate action required. [Key answer]

D Withdraw the fuel assembly and contact Reactor Engineering for guidance.

Facility Comments:

Page 7 of 11

Post Exam Comments Millstone Unit 2 March 2005 Initial License Exam Comments:

Choice C (the correct answer) is an acceptable answer for an anticipated count rate multiplication due to the loading of a new fuel assembly in a location adjacent to the CH B Wide Range detector. Since no information is supplied in the stem of the question as to the refueling method (e.g. Full Core Reload) it is reasonable to expect this change in some instances.

Choice B is also an acceptable response. The stem of the question provides no additional information regarding the method of refueling and status of refueling. Assumptions could be made as to the refueling method (e.g. Fuel Shuffle) and previous data trends of a 1/M plot and/or count rate changes.

Historical data shows that during a fuel shuffle the 1/M value rarely dips below (.8). Using the provided count rate data from the stem of this question l/M values ( C&,iaJCR,, ) are as follows; CH A (.95). CH B

(.56 almost a doubling), CH C (.84), CH D (.83). If an assumption is made that initial count rates at the start of this fuel shuffle was even lower, then its effect on a 1/M plot would be greater and a doubling may be evident. If the operator believes that an unanticipated count rate multiplication has occurred hdshe is compelled by OP 2209A to commence an emergency boration.

Recommendation:

Accept Choice C and Choice B Justification:

Choice B correctly describes the required operator action of OP 2209A and is a conservative response to a situation that required judgement.

4.5.15 IF, ant any time, unanticipated count rate multiplication, (i.e., doubling), is indicated, PERFORM the following:

a. SUSPEND refuel operations.

b. Refer To AOP 2558, Emergency Boration and PERFORM applicable actions to initiate boration to RCS.

c. Immediately NOTIFY Reactor Engineering and SM.

d. REQUEST evaluation be completed prior to restart of fuel handling activities.

e. INITIATE CR.

NRC Resoonse:

Facility recommendation for multiple correct answers is NOT ACCEPTED.

As stated in the question and the facility comment, the procedural requirement is to commence an emergency boration if an unanticipated doubling of count rate occurs. A core map was included with this question for the applicants to use as an attached reference: Using the core map, applicants could determine from information in the question stem that the inserted assembly was adjacent to the wide range logarithmic power instrument that is showing increased counts. The count rate increase indicated on this channel has not doubled from the initial readings given. The other channels are indicating only very slight rise in counts. For the given conditions, emergency boration is not required. In addition, if we were to accept the facility argument the B was appropriate, then answer A would be correct as well, a third correct answer requiring deletion of this question rather than acceptance of a second answer.

Applicants are cautioned at the start of the written exam per NUREG 1021, Appendix E, to not make assumptions regarding conditions that are not specified in the question unless they occur as a consequence Page 8 of 11

Post Exam Comments Millstone Unit 2 March 2005 Initial License Exam of other conditions that are stated in the question". Presuming that initial counts were even lower at the start of refueling requires the applicant to make assumptions not stated in the question.

References:

- OP-2209A, "Refueling Operations", Revision 24 Page 9 of 11

Post Exam Comments Millstone Unit 2 March 2005 Initial License Exam

6. JPM A2SRO A step identified as critical in this SRO JPM required applicants to specify the procedure that contains the necessary post-maintenance testing guidance.

Facilitv Comments:

Justification for accepting 2604AO and 2308x1 1 as a correct answer to JPM-A2SRO SRO AWO Acceptance:

Millstone Unit 2 is developing a new set of procedures designed to simplify component maintenance and retesting. These new procedures, called Maintenance Operating Procedures (MOPs), are designated as 2300X. The MOPs include the steps of the 2600 procedure and additional sections for venting, draining and tagging for the component in question. When a MOP is approved, it becomes the preferred post-maintenance procedure and replaces the previously used 2600.

At the time the SRO candidates were trained and when this JPM was developed, the MOPs did not exist. At the time of the NRC Exam, a MOP was approved for the A HFSI Pp. I believe this is the only MOP that is approved for use at this time.

Due to the present state of transition from the 2600 procedures to the 2300X procedures, Millstone Unit 2 requests both 2604AO and 2308x1 1 be considered correct.

NRC Response:

Recommendation to allow both 2604AO and 2308x1 I as correct procedure references is ACCEPTED.

The 2300 procedure covering post-maintenance testing of HPSI Pump A (relating to this JPM) has been issued, and is the correct procedure for the evolution. However, the license applicants were trained to use the 2600 plant equipment TS surveillance procedures as post-maintenance test procedures. An applicant, not yet trained on recent procedure changes, would be expected to identify the 2600 procedure as the correct procedure. The 2600 procedures will continue to exist and are used for conduct of periodic scheduled surveillances. The 2300 series procedures provide more comprehensive guidance, including steps for drain, refill and testing to restore operability.

References:

- 2604A0, HPSI PUMP INSERVICE TESTING, 1,750 PSIA, FACILITY l, Revision OOO-00

- 2308x1 1, A HFSI PUMP MAINTENANCE, Revision 000-01 Page 10 of 11

Post Exam Comments Millstone Unit 2 March 2005 Initial License Exam

7. JPM A l R O Several steps identified as critical in this SRO JPM appeared to have incorrect error bands.

Facilitv Comments:

During grading of this JPM, the NRC noted several applicants who obtained the correct end result despite errors in intermediate steps identified as JPM critical steps. The NRC grader determined that there were errors in the answer key for this JPM and requested a corrected key.

NRC Response:

The NRC verified that the resubmitted key contained the corrections identified by the NRC grader.

8. JPM A2RO Facilitv Comments:

Tagging of the chiller compressor breaker in this JPM should not have been a critical step. The compressor is interlocked with the associated chilled water pump, so tagging the chill water pump prevents the compressor from starting.

NRC Response:

The facility requested change to the answer key is ACCEPTED.

The NRC reviewed the chilled water system OP 233OC. which addresses system maintenance and does not require the chiller compressors to be tagged under the conditions in this JPM. The NRC also reviewed the tagging procedure WC-2 and noted that this procedure does not preclude dependence on interlocks for equipment protection. No procedural basis was found for requiring this breaker to be tagged.

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