Text

Initial Exam 2010-301 Draft SRO Written Exam
	ML102650109
Person / Time
Site:	Mcguire, McGuire
Issue date:	08/30/2010
From:	; NRC/RGN-II
To:	; Duke Energy Carolinas, Duke Power Co
References
	50-369/10-301, 50-370/10-301 50-369/10-301, 50-370/10-301
	Download: ML102650109 (166)
	v • d • e

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2010 MNS SRO NRC Examination QUESTION 76 KA_desc SYSOO3 SYSOO3 GENERIC Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 /45.12) 2.1.20 Given the following conditions on Unit 1:

The unit was initially at 100% RTP

#1 Seal Leakoff on 1A NC pump indicates 6.5 GPM

AP-08 (MaUunction of NC Pump) Case I (NC Pump Seal or Pump Lower Bearing Malfunction has been implemented

The crew has reached the steps in AP-08 to trip the Reactor and stop the 1A NC pump In accordance with AP-08, Enclosure 2 (NC Pump Post Trip Actions For #1 Seal Failure) must be performed within 3-5 minutes after stopping the 1A NC pump to prevent (1) . The requirement to perform these actions is applicable (2)

Which ONE (1) of the following completes the statement above?

A. 1. damage to the IA NC pump #2 & #3 seals

2. only while AP-08 is in effect B. 1. damage to the 1A NC pump #2 & #3 seals

2. even after transition from AP-08 to E-0 C. 1. the VCT from exceeding design temperature limits

2. only while AP-08 is in effect D. 1. the VCT from exceeding design temperature limits

2. even after transition from AP-08 to E-0 Friday, May 14, 2010 Page 201 of 275

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2010 MNS SRO NRC Examination QUESTION 76 General Discussion In accordance with AP-08, after the Reactor is tripped and the NC pump is stopped the seal return valve for the AFFECTED NC pump (1NV-34A) must be closed within 3-5 minutes. This action is contained in Enclosure 2 (NC Pump Post Trip Actions For #1 Seal Failure). In accordance with the AP-08 Background Document, the seal return line must be isolated to prevent damage to the #2 and #3 seals due to high temperature water flowing past the seals.

Per AP-08, the requirement to close 1NV-34A within 3-5 minutes after stopping the pump is applicable even after transition to the EPs.

Answer A Discussion INCORRECT. See explanation above.

PLAUSIBLE: Part 1 is correct.

Part 2 is plausible because conditional steps in APs are typically no longer applicable when transition is made to the EPs. In this particular case the applicant must recall the caution from AP-08 that states the post pump trip actions in the AP-08 enclosure are applicable even after transition to the EPs to arrive at the correct answer.

Answer B Discussion CORRECT. See explanation above.

Answer C Discussion INCORRECT. See explanation above.

PLAUSIBLE: Part 1 is plausible because this is the basis for closing 1NV-94AC and 95B (NC Pumps Seal Return Cont Isolations) in AP-08 when both seal injection and thermal barrier cooling are lost.

Part 2 is plausible because conditional steps in APs are typically no longer applicable when transition is made to the EPs. In this particular case the applicant must recall the caution from AP-08 that states the post pump trip actions in the AP-08 enclosure are applicable even after transition to the EPs to arrive at the correct answer.

\nswer D Discussion 4

INCORRECT. See explanation above.

PLAUSIBLE: Part 1 is plausible because this is the basis for closing 1NV-94AC and 95B (NC Pumps Seal Return Cont Isolations) in AP-08 when both seal injection and thermal barrier cooling are lost.

Part 2 is correct. -

Basis for meeting the KA The applicant demonstrates the ability to interpret procedure steps by demonstrating a knowledge of basis for performing the Post Pump Trip Actions of Enclosure 2 (specifically closing the NC pump seal return valves within 3-5 minutes). The applicant demonstrates the ability to execute procedure steps by demonstrating the knowledge that the AP procedure steps must be performed even after transition to the EPs.

Basis for Hi Cog -______

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions (Rev 1 dated 03/11/2010) under the Screening Criteria for question linked to 10CFR55.43(b)(5) (Assessment and Selection of Procedures):

1) The question can NOT be answered by knowing systems knowledge alone. The basis for closing the seal return isolation valve for the affected pump within 3-5 minutes is not covered by the NCP system lesson plan. Therefore, this is not systems level knowledge.

2) The question can NOT be answered by knowing immediate Operator actions.

3) The question can NOT be answered by knowing AOP or EOP entry conditions.

4) The question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of the procedure.

5) The question requires the applicant to recall procedure content from AP-08 (i.e. that the Post Pump Trip Actions must still be performed even after transition to the EPs). Additionally, the applicant must recall why the procedure steps must be performed from the AP-08 Background Document.

ZJob Level Cognitive Level QuestionType Question Source SRO Memory NEW Friday, May 14, 2010 Page 202 of 275

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2010 MNS SRO NRC Examination QUESTION 76

,1 Jveiopment References udent References sson Objective:

N/A

References:

1) AP-08, Malfunction of NC Pump Document KA KA_desc SOO3 JYSOO3 GENERIC EAbility to interpret and execute procedure steps. (CFR: 41.10 / 43.5 / 45.12) 2.1.20 401-9 Comments: RemarkslStatus Friday, May 14, 2010 Page 203 of 275

Question 76

References:

From AP-08:

I ACflCU/EflCTED EEPONSE NOT DatA:N:D I

3. Check if scal cooling available to Perform the following:

affected pump Cl USE me of th follrxhg c Seal injection est3blished U\ornial 0 5SF supplvi

1NV-91AC (NC Rjmps Seal Ret Cont Inside sd)

OR OR a XC Lu .Fieiiiial bdIri*I &bIisIwJ.

. 1NV-95D (NC Punis Seal Ret Court Outside Isoli.

I

_b. IFLTANYTIMEseiI :oolirg I; restored. THEN cbser.e Nole prior tD Step 7 and GO TO Step 7.

c. RETURN TO SteD 2.

NOTC Lp to 24 ioursof NC r&irnp operatioi rna\ be required before seals seat aid operate normally after seal maintenance cr startup.

7. Check any NC pump mrber 1 se& Perform the following:

leakolt GRE4TER Tt-IkN OR EQUAL TO 6GPM.

NUIb CP!l/ii62UWUUib (Chem cal and Vclu me Cuiiio SysLrr I

Cha-ging. Endoeure 4 10 (Maintaining NC Rsvp Seal LeakcIf i g ye; guidarce on a:tions usec to charge seal leakoff flow.

a. IF seal leakoffslcwl, goirg up, THEN contact s:aton managenient for further guidance.

b. Continue to nioni:or NC lxJrrp seal leakoff flew.

c. IFATANYflMFsnI ankoff flow goes up to 6 M. THEN GD TO Step 6.

_d. cQIQaep9

MNS MALFUNCTION OF NC PUMP PAGE NO.

AP1I /AJ5500/08 Case I 6 of 24 UNIT 1 NC Pump Seal or Pump Lower Bearing Malfunction Rev. 12 ACTICN/EXPECTED RESPONSE RESPONSE NOT OBTAINEO

8. Stop affected NC pump as follows:

a. fflA or B NC pump is the affected pump, THEN CLOSE associated spray valve:

. INC-27C (A NC Loop PZR Spray Control)

. 1NC-29C (B NC Loop PZR Spray Control).

CAUTION Enclosure 2 (NC Pump Post Trip Actions For #1 Seal Failure) contains actions that must be performed between 3 and 5 minutes after stopping NC pump. This enclosure must be performed even after transition to EPs.

b. Have any available RO perform Enclosure 2 (NC Pump Post Trip Actions For #1 Seal Failure) as crew performs the following steps.

MNS MAl Ft JNCTION OF NC Pt iMP PAGF NO API 1 1N5509?08 Enclosure 2- Page 1 of 2 UNIT I NC Pump Post Trip Actions For #1 Seal Failure e\. 12 R

A:ncn/EXPE:rEz RESIONSE RESPONSE NOT OaTADCED CAUTiON Failure ofNuniber2 and 3 seals may occur unless the affected NC pump seal return valve is closed between 3 minutes and 5 minutes after slopping puiiip. This enclosure must be completed even after transiion to EPs.

1. Record time of NC pump shutdown:

2. Check if seal cooing available to Perform the following:

affected rump:

a. CLOSE the following:

. Seal injection esiablished {Nomml r 5SF Sinplv) . 1 NV-94AC (NC Pumps Seal Ret Cont Inside Isol) fl OR

. 1NV--Jbb (NC Pumps Seal Ret Cont

. KC to fremml barier estaolishecL Outside Isoli.

b. Exit this enclosure.

3. Check an, NC pump number I sea. GO TO Step 5.

leakoffflcw GREATER THAN OR EQUAL TO 6 GPM.

4. Waintain seal injection flow greater than 9 GPM to affected pump(s).

r\4NS 1ALFUNCTlON OF NC PUMP PAGE NO AP/1 A!55OO/O8 24 of 24 Enclosure 2 - Page 2 of 2 IJNIT 1 NC Pump Post Trip Actions For #1 Seal Failure Rev 12 CTICN,/EXPECTED RSPON RESPONSE O1? OBTAINED

5. WHEN affected NC pump has been off 3 minutes THEN immediately perform the following:

a CLOSE affected NC pump seal return valvG:

. 1 NV-34A (A NC Pump Seal Return Isol)

I NV-50B (B NC Pump Seal Return Isol)

. I NV-36A (C NC Pump Seal Return Isol)

1 NV-82 (D NC Pump Seal Return Isol).

b. OPEN Ji of the following valves:

. OPEN IKC-394A (A NC Pump Therni Bar OtIt).

OPEN 1KC-345A (C NC Purnr Therm Bar Otlt).

. OPEN 1KC-364B (B NC Pump Therm Bar OtIt).

OPEN 1KC-413B (D NC Pump Therm Bar Otlt)

From AP-08 Background Document:

APII and 21A155001008 (Malfunction of NC Pump)

CASE I STEP 6:

PURPOSE:

Prevent hot NC Pump seal return flow from going to the VCT and prevent transition to the steps that may close inctividual pump seal return valves if no seal cooling exists.

DISCUSSION:

With no seal injection coincident with no thermal barrier cooling, this step closes NV-94 and 95 (NC Pump seal return containment isolations). This will force the hot *1 seal leak-off flow to the PRT and prevent the VCT from exceeding design temperature limits (150 F). The RNO of this step will also prevent transition to the next several steps dealing with specific seal failures, and back to monitoring for NCP trip criteria. This has the benefit of skipping the steps that would close the individual seal return isolation valves. The individual seal return isolation valves need to remain open for loss of all seal cooling events. The following is an excerpt from DW-94-O 11:

Isolation of the 1 seal leakoff line during a loss of all seal cooling event would force the #2 NCP seal into the high pressure mode of operation at high temperature. This is beyond the design basis of the 2 seal, and the response of the 2 seal to high pressure operation without cooling is unknown. The analysis performed for the extended loss of ac power in WCAP 10541, Rev. 2, identifies that lhe high temperature two-phase flow through the seal system and

1 seal leakoff line increases the pressure in the #1 leakoff cavity. The increased leakoff cavity pressure tends to decrease the separation of the #1 seal faces which tend to reduce the leakage. The combination of competing effects results in higher leakage rates, but leakage rates that are self-limiting. Higher flow increases the system back pressure which reduces the separation of the faces, reducing the flow. Therefore, keeping the #1 seal leakoff line open will provide the benefit of minimizing the leakage, while closing the leakoff line could result in catastrophic failure of the #2 seal and actually increase leakage.

In summary, closing NV-94 & 85 keeps individual flowpath open for #1 seal leakoff (through relief to PRT), but isolates it from the VCT, Note: this AP does not address restoring a loss of seal injection or thermal barrier cooling.

REFERENCES:

DW-94-01 I page 6 of 7 WCAP-10541. Rev.2 Page 7 of 39 Rev 2

APII and 21A155001008 (Malfunction of NC Pump)

CASE I STEP 8:

PURPOSE:

Provkle the direction for stopping the affected NC Pump.

DISCUSSION:

Closing a spray valve for A or B pump ensures Pzr pressure control is maintained. The operator is cued to close the spray valve in anticipation of losing the motive force for flow through the valve. The operator is expected to take whatever other compensatory action is required to stabilize Pressurizer pressure (operating heaters or other spray flow).

Between three and five minutes after the pump has been tripped, the affected pumps seal return isolation is closed. Waiting three minutes ensures the pump has stopped rotating. The

2 seal has a softer seat than the #1 seal, and if rotating while exposed to the potential debris from the failed #1 seal, it could experience premature failure. Closing it in less than 5 minutes minimizes the time the #2 seal is exposed to high temperature fluid conditions.

Closing the seal return valve within 3 to 5 minutes during a #1 seal failure event does not meet the criteria of high PRA values as determined by the Severe Accident Analysis Group. For McGuire, this corresponds to a Risk Achievement Worth (RAW) greater than or equal to 1 .04.

PIP M-03-1992 documents the events that meet these criteria. As such this action is not a McGuire time critical action, but is a management expectation and prudent action to prevent damage to the #2 and #3 seals. PIP M-07-031 0 ACA#4 documents the removal of this action from McGuires time critical action list.

After the affected seal return is closed, the thermal barrier outlet valve is opened, if necessary.

This Is after the previous step to ensure its after any perturbations that would close the valve.

An available RO is designated to perform Enclosure 2 (NC Pump Post Trip Actions for #1 Seal Failure). This could be an extra RO if available. If one is not available, this could be the BOP.

It has to be someone. The use of an enclosure facilitates the designated person completing the enclosure actions while minimizing the interaction with the crew. The enclosure can be handed off to the designated person while the crew can focus on just E-0. With an enclosure, additional communications between the RO and the SRO arent required to complete these actions.

If in Mode 1 or 2, the guidance for stopping a NC Pump includes: tripping the reactor, waiting for reactor power to decay below 5%, tripping the affected NC pump, and transition to E-O. In Mode 3, 4, or 5, this is not necessary. This is why the two steps for stopping an NCP are written differently.

Guidance is given to wait until reactor power is less than 5% before stopping the NC pump. This will ensure the NC pump will provide adequate flow/core cooling until reactor power is sufficiently low enough to preclude a challenge to fuel integrity.

PagelOof39 Rev2

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2010 MNS SRO NRC Examination QUESTION 77 fl57j KA_desc SYSOO5 Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 A2.02 /43.5 /45.3 / 45. 13)ElPressure transient protection during cold shutdown Given the following conditions on Unit 1:

The unit is in Mode 5

NC system temperature is currently 112°F

lAND Train is in service

A special test procedure is to be run which requires BOTH NI pumps to be run in parallel and aligned to inject into the NC system.

Which ONE (1) of the following describes the requirements per Tech Spec 3.4.12 (LTOP) Bases?

A. Secure two PORVs open with associated block valves open and power removed.

This action protects against brittle fracture due to pressurized thermal shock of the reactor vessel.

B. Secure two PORVs open with associated block valves open and power removed.

This action protects against brittle fracture due to cold overpressure of the reactor vessel.

C. Establish an RCS vent of 2.75 square inches and verify at least ONE Operable PZR PORV.

This action protects against brittle fracture due to pressurized thermal shock of the reactor vessel.

D. Establish an RCS vent of> 2.75 square inches and verify at least ONE Operable PZR PORV.

This action protects against brittle fracture due to cold overpressure of the reactor vessel.

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2010 MNS SRO NRC Examination QUESTION 77 eneraI Discussion Based on the conditions given, the applicant is placed in a condition that if the test is to be run, certain conditions must be met to satisf the LTOP vent path requirements.

One method of meeting the vent path requirements is to establish an adequate vent path prior to starting the test. For this case, securing open two Pressurizer PORVs or establishing a vent path of greater than or equal to 4.5 will meet those requirements.

Another method of meeting the vent path requirements is establish a RCS vent path of> 2.75 AND two Operable PORVs.

The second part of the question deals with the basis of LTOP which is the protection of the reactor vessel from brittle fracture at lower temperatures.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE: Part 1 is correct and therefore plausible.

Part 2 is plausible because pressurized thermal shock is a low temperature brittle fracture event but is predicated by a rapid overcooling event which sets up a temperature gradient across the reactor vessel. One of the actions which would allow for this test is the verification of an operable RHR suction relief and that NCS temperature is greater than 74° and Cool Down rate <20 Deglhr. The applicant could misinterpret this as preventing a PTS type event.

Answer B Discussion CORRECT: See explanation above Answer C Discussion I1JCORRECT: See explanation above.

PLAUSIBLE: Part 1 is plausible because it is partially correct in that the actions stated would meet LTOP requirements but an additional perable PZR PORV would required.

Part 2 is plausible because pressurized thermal shock is a low temperature brittle fracture event but is predicated by a rapid overcooling event which sets up a temperature gradient across the reactor vessel. One of the actions which would allow for this test is the verification of an operable RHR suction relief and that NCS temperature is greater than 74° and Cool Down rate <20 Deg/hr. The applicant could misinterpret this as preventing a PTS type event.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE: Part 1 is plausible because it is partially correct in that the actions stated would meet LTOP requirements but an additional operable PZR PORV would required.

Part 2 is correct and therefore plausible.

Basis for meeting the KA -________________________________

The KA is matched because an operation is about to occur (operation of two NI pumps at the same time) that would result in a pressure transient in the RCS during cold shutdown. The applicant is asked to predict the possible impacts (Brittle fracture) and determine the requirements to mitigate the possible consequences of the proposed test. How this test will impact the LTOP system vent path requirements (Pressure transient protection during cold shutdown). The applicant must determine the actions required by Tech Specs (use procedures to control) that will allow both NI pumps to be run simultaneously.

Basis for Hi Cog This is a higher cognitive level question because it requires multiple mental steps. The applicant must first analyze the given information to determine that the vent path requirements have changed from 2.75 to 4.5. The applicant must then recall from memory all combinations of equiment that would meet the 4.5 vent path requirement.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions (Rev 1 dated 03/11/2010) under the Screening Criteria for question linked to 10CFR55.43(b)(2) (Tech Specs):

1) It can NOT be answered solely by knowing < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs.

I It can NOT be answered solely by knowing the LCO/TRM information listed above-the-line.

) It can NOT be answered by knowing the Tech Spec Safety Limits or their bases

4) It requires the applicant to have detailed knowledge of Tech Spec 3.4.12 vent path requirements and information from the TS 3.4.12 Basis Document to determine the correct answer.

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2010 MNS SRO NRC Examination QUESTION 77 2577 Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided Learning Objectives:

1) PS-NC #24

References:

Tech Sped 3.4.12 Basis KA KA_desc SYSOO5 Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 A2.02 / 43.5/45.3 / 45.13)EPressure transient protection during cold shutdown 401-9 Comments: RemarkslStatus Friday, May 14, 2010 Page 206 of 275

Question 77

References:

From OP-MC-PS-NC Objectives OBJECTIVES NNLLL OBJECTIVE L L P P 0 OORSR ROO 20 Describe how NCS temperature, pressure, flow and Pzr level X X X are measured and indicated.

21 Describe the operation and indication readout of the following X X X X NCS level instrumentation:

. Ultrasonic level detection

. WR level

. NRlevel

. Sightglass 22 State the nominal values for NC System pressure, Th, Tc, X X X X Tave, Pzr temperature for Hot Zero Power and Hot Full Power.

23 Given a Limit and/or Precaution associated with the NC X X X X System, discuss its basis and when it applies.

24 Concerning the Technical Specifications related to the NC X X X System: x x x

. Given the LCO title, state the LCD ( including any x x x COLR values) and applicability.

x x x

. For any LCOs that have action required within one x

hour, state the action.

. Given a set of parameter values or system conditions, determine if any Tech Spec LCOs is(are) not met and any actions(s) required within one hour.

. Given a set of parameter values or system conditions and the appropriate Tech Spec, determine required action(s).

. Discuss the bases for a given Tech. Spec. LCO or Safety Limit.

SRO ONLY

From TS 3.4.12 Basis:

LTOP System B 34.12 OASES APPLICABILITY (continued) the pressurizer safety waIves that provide overpressure protection during MODES 1. 2. and 3, and MODE 4 above 300DF.

Low temp erature overpressure prevention is most critical during shutdown when the RCS is water solid, and a mass or heat input transient can cause a very rapid increase in RCS pressure when little or no time ailos operator action to mitigate the event.

The Applicability is modified by a Note stating that accumulator isolation is only required when The accumulator pressure is more than or at the maximum RCS pressure for the existing temperature, as allowed by the PiT limit curves. This Mote permits the accumulator discharge isolation valve Surveillance to be performed only under these pressure and temperature conditions.

ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable LTOP system. There is an increased risk associated with entering MODE 4 from MODE 5 with LTOP inoperable and the provisions of LCO 3.0.4 .b.

which allow entr into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

A.i, All. A.2.2.i. A22.2. A.3, A.4, A.5.1, andA.5.2 centrtfuqal charqinq pumps, or a combination of each, capable of injecting into the P08, ROS overpressurization is possible.

To immediately initiate action to restore restricted coolant input capability to the RCS reflects the urgency of removing the P05 from this condition.

Two pumps may be capable of injecting into the RCS provided the RHR suction relief valve is OPERABLE with:

Cowed answer. one option to a]Iow the 1. RCS cold leg temperature > 174°F (Unit 1), or ter would be to 2. RCS cold leg temperature> 89°F (Unit 2). or ecuie open 2 3. ROS cold leg temperature > 74°F and cooldosn rate c20°Fihr (Unit 1),

POKVz or

4. RCS cold leg temperature > 74°F and cooldown rate 60°Fihr (Unit 2),

I

6. a F..cSvenfl r 4.5 square inches, or McGuire Units 1 arid 2 B 3.4.12-7 Revision No. 102

BASES ACTIONS (continued)

3 os vent of 2.75 sci..are inches arid two CPERABLE RS *;ent shal not be .tie of t:,et..cc OPERABLE PORVs Thit provides the For cases where no reactor coolant pumps are in operation, RCS cold leg plauzibthty for temperature limits are to be met by monitoring of BOTH the VV R Cold Leg dttcter: A and B. temperatures and Residusl Heat Demoval Heat Exchaner discharo,e temperature. rTWEThh PORYS and block vaive seured operMt r RGb vent of 4.5 square inches. there are no credible sinqaiLres limit the flow relief capachv. For the RH re Ic :aliero be OP RABLE the RIHR suction so atkr valves must be a.neflet S ctt45Qiq teru:ith he . The RHR suction relief valves are spring loaded, bellows type water relief valves with pressure tolerances and accumulation limits established by Section III of the American Society of Mechanical Engineers (ASME) Code (Ref. 3)for Class 2 relief valves.

Required Action & 1 is modified by a Note that permits two centrifugal charging pumps capable of RCS injection for 15 minutes to allow for pump swaps B:l, 0.1. and C.2 An unisolated accumulator requires isolation within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . This is ont?

required when the accumulator pressure is at or more than the maximum RCS pressure for the existing temperature allowed by the PiT limit curves.

If isolation is needed and cannot be accomplished in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , Required Action C. I and Required Action C.2 provide two options, either of tiich must be performed in the ne 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . By increasing the RCS temperature to> 3C0°F, an accumulator pressure of 639 psig cannot exceed the LTOP limits lithe accuniulators are fully injected.

Depressurizing the accumulators below the LTOP limit also gives this protection.

The Completion Tines are based on operating experience that these activifies can be accomplished in these time periods and on engineerir evaluations indicating that an event requiring LTOP is not likely in the allowed times.

0.1 In MODE 4 when any RCS cold leg temperature is 300° F. with one PORV inoperable, the PORV must be restored to OPERABLE status McGuire Units 1 and 2 B 3.4.12-8 Revision No. 102

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2010 MNS SRO NRC Examination QUESTION 78 25 KA_desc SYSO61 Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions,

- use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5/45.3 A2.06 / 45.13)ElBack leakage of MFW Given the following conditions on Unit 1:

The unit is at 12% RTP preparing to roll the main turbine

M1A1276 (UI CA Temp at Chk Vlv I CA-37) alarms on the OAC

1CA-37 (#1 TD CA to SIG D)

Based on the above conditions:

1. What method would FIRST be used to reduce the temperature at the check valve?

2. How would this action affect the operability of the TD CA Pump?

A. 1. Close 1 CA-36 AB (Ui TD CA Pump Disch to 1 D S/G Control) and monitor temperature for 15 mm.

2. The U-i TD CA Pump remains OPERABLE.

B. 1. Close 1 CA-36 AB (Ui TD CA Pump Disch to I D S/G Control) and monitor temperature for 15 mm.

2. The U-i TD CA Pump shall be declared INOPERABLE.

C. 1. Close 1 CA-38B (UI TD CA Pump Disch to 1 D S/G Isol) and start the TD CA pump aligned for recirculation to the UST.

2. The U-I TD CA Pump remains OPERABLE.

D. I. Close 1 CA-38B (UI TD CA Pump Disch to ID SIC Isol) and start the TD CA pump aligned for recirculation to the UST.

2. The U-i TD CA Pump shall be declared INOPERABLE.

Friday, May 14, 2010 Page 207 of 275

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 78 eneraI Discussion The consequence of the situation described would be overheating of the TD CA pump discharge piping which could lead to voiding and ultimately steam binding associated with this pump. The correct response to the alarm associated with OAC point M1A1276 is to reduce CA system piping temperature per OPI1/A/62501002. Enclosure 4.4 of this procedure directs the operators to first close the control valve on the affected line, which in this case would be 1CA-36AB or the D SIG. If this is unsuccessful, then the pump is run in recirc to cool the discharge line but all of the remaining motor operated control valves would have to be closed first and this would only be done if the closure of the single control valve was not successful. The stem of the question asked for the FIRST action.

The operability of the TD CA pump is affected both by the closure of the Air Operated flow control valves (ICA-36 AB). Above 10% RTP,

[ing this valve renders the pump inoperable.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE: Part I is correct and therefore plausible.

Part 2 is plausible because it the unit was below 10% RTP the action of closing the control valve would not affect the operability of the associated AFW pump. This answer is plausible because it is possible to close this valve with the unit at power without affecting operability just not at the given power level.

Answer B Discussion iiRRECT: See explanation above.

Answer C Discussion iORRECT: See explanation above.

1 PLAUSIBLE: Part 1 of the answer is plausible because this action is correct but the stem of the question asked for what method would be used First. The method described would only be employed if the closure of the control isolation was not successful but since it is a possible strategy, it is plausible.

Part 2 is plausible because it the unit was below 10% RTP the action of closing the control valve would not affect the operability of the

sociated AFW pump. This answer is plausible because it is possible to close this valve with the unit at power without affecting operability just not at the given power level.

Answer D Discussion INCORRECT: See explanation above, PLAUSIBLE: Part I of the answer is plausible because this action is correct but the stem of the question asked for what method would be used First. The method described would only be employed if the closure of the control isolation was not successful but since it is a possible strategy, it is plausible.

2 is correct and therefore plausible.

Basis for meeting the KA of this question matches the a part of the KA regarding predict the impact of the following malfunctions on the AFW. The impact is whether the TD CA pump will remain operable.

Part I of this question matches the part b of the KA regarding using procedures to correct, control, or mitigate the consequences. The

[pedure in this case is OP/l/A!6250/002, Auxiliary Feedwater System, Enclosure 4.5, Reducing Turbine Driven CA Pump Piping Temperature.

Basis for Hi Cog This question is Hi Cog because the applicant must evaluate a given set of conditions and through a multipart mental process, determine the required actions based on these conditions. The applicant must futher evaluate the impact of the actions to address the high temperature on the operability of the associated AFW pump.

Basis for SRO only Part 1 of the question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev I dated 03/11/2010 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures):

1) The question can NOT be answered solely by knowing systems knowledge. Either of these methods can be used procedurally to cool the TD CA pump piping. Check valve leakage is discussed in the systems lesson plan and the methods to cooldown the TD CA pump are mentioned in qeneral terms (i.e. close the discharge valve or start the pump). However, the applicant must have detailed knowledge of the OP to scrimminate which method is used FIRST. Since this is an infrequently performed evolution, the actions in the procedure are directed by the R SRO and not handed off to an RO.

2) The question can NOT be answered by knowing immediate operator actions. None of the actions described are immediate actions.

3) The question can NOT be answered solely by knowing entry conditions for AOP or direct entry conditions for EOPs. These are detailed procedure steps from an infrequently performed OP.

Friday, May 14, 2010 Page 208 of 275

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4) The question can NOT be answered solely by knowing the purpose, overal sequence of events, or overall mitigative strategy of the procedure.

his is detailed knowledge of procedure step sequence not sequence of events within the procedure.

The question requires detailed knowledge of procedure content. Therefore, it is SRO knowledge.

Part 2 of the question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for screening questions linked to 10CFR55.43(b)(2) (Tech Specs):

1) This question can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs

2) This question can NOT be answered by knowing information listed above-the-line.

3) This question can NOT be answered by knowing the TS Safety Limits.

4) This question required the applicant to analyze the given conditions and make the determination that the TD CA pump is inoperable. The applicant must then recall from memory that the unit can not enter MODE 1 with the TD CA pump INOPERABLE.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided TS 3.7.5 OP/1/A16250/002 Auxiliary Feedwater System KA KA_desc SYSO61 Ability to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 A2.06 /45.13)LlBackleakageofMFW (NO1-9 Comments: RemarksiStatus Friday, May 14, 2010 Page 209 of 275

Question 78

References:

3.7.5 Auxiliary Feedwater (AFW) System LCO 3.7.5 Three AFW trains shall be OPERABLE.

NOTE Only one AFW train, which includes a motor driven pump, is required to be OPERABLE in MODE 4.

APPLICABILITY: MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.

ACTIONS NOTE LCO 3.0.4.b is not applicable when entering MODE 1.

CONDITION REQUIRED ACTION COMPLETION TIME A. One steam supply to A.1 Restore steam supply to 7 days turbine driven AFW OPERABLE status.

pump inoperable. AND 10 days from discovery of failure to meet the LCO B. One AFW train B. 1 Restore AFW train to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable in MODE 1, 2 OPERABLE status.

or 3 for reasons other AND than Condition A.

10 days from discovery of failure to meet the LCO (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.5.1 Not applicable to autol IYL !1it1i1 1 POWER is <10% RTF Verify each AFW manual, power operated, and automatic 31 days valve in each water flow path, and in both steam supply flow paths to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.7.5.2 Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after 900 psig in the steam generator.

Verify the developed head of each AFW pump at the flow In accordance test point is greater than or equal to the required with the Inservice developed head. Testing Program SR 3.7.5.3 NOTE Not applicable in MODE 4 when steam generator is relied upon for heat removal.

Verify each AFW automatic valve that is not locked, 18 months sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

From 0P111A162501002, Auxiliary Feedwater System:

Enclosure 45 OP/1JAj6250/002 Reducing Turbine Difren CA Pump Piping Page 1 of 3 Temperature

1. Limits and Precautions None

2. Initial Conditions 2.1 8W System isolated from S/Os per OP/1/A/6l0OSO-5A (B. C. 1)) (Draining S/Cl 1A. lB. LC. ID).

3. Procedure Q 3.1 Evaluate all outstanding R.&Rs that may impact performance of this procedure.

3.2 Declare #1 ID CA Pump inoperable.

sP.o 3.3 Close control valve on affected lines:

1CA-MAB (Ui TD CA Pump Disch to 1A S/Cl Control) cv

ICA-52AB (UI IT) CA Pump Disch to 18 S/Cl Control) cv

1CA-4SAB (Ui IT) CA Pump Disch to 1C S/Cl Control) cv

1CA-36AB (Ui ID CA Pump Disch to 1D S/Cl Control) cv Q 3.4 Monitor temperature for 15 30 minutes.

15 IF temperatures remain high after 15 30 minutes. close isolation valve on affected lines:

1CA-6AC (Ui ID CA Pump Disch to lAS/Cl Isol) cv

ICA-54AC (Ul TD CA Pump Disch to 18 S/Cl Isol) cv

ICA-50B (1.11 ID CA Pump Disch to 1C S/Cl Isol) cv

ICA-38B (Ui ID CA Pump Disch tolD S/Cl Isol) cv Unit 1

EncLosure 4.5 OPL\.6250.002 Reducing Turbine Drhen CA Pump Piping Page 2 of 3 Temperature NOTE: When opening valves 1CA-36. 48. 52. and 64 toni the local panel, the controller needs to be opened 4 5 more turns once 1OO0. is reached to minimize the amount that the valves drift close and back open upon returning controller back to control room (A-Remote).

3.6 AFTER temperatures have returned to normal, ensure open:

ICA-64AB (Ui TD CA Pump Disch to 1A St Control) cv

1CA-S2AB (UI ID CA Pump Ditch to lB St Control) cv

ICA-4SAB (UI ID CA Pump Disdh to 1C 510 Control) cv

ICA-36AB (Ui ID CA Pump Disch to iD St Control) cv

1CA-66AC (Ui TD CA Pump Disch to 1A St isol) cv

ICA-S4AC (Ui TD CA Pump Disch to lB 570 Isel) cv

,. ICA-50B(Ul TDCAPumDIschto1CS/OThol) cv

ICA-38B (Ui ID CA Pump Ditch to 1D SiC Isol) cv 3.7 Check the following stable:

D M1A1439 (UI CA Temp at Cit iv 1CA-65) o M1A1421 (UI CA Temp at Cli Vlv 1CA-53) 0 MlAl2g4 (UI CA Temp at Cli Vlv 1CA-49) 0 M1AI 276 (Ul CA Temp at Cli Vlv 1CA-3 7)

[nit 1

Enclosure 4.5 . 11462501002 Reducing Turbine Driven CA Pump Piping Page 3 of 3 Temperature 3.8 II increasing temperatures indicates check valve leak by perform the following:

38.1 Notify System Engineer.

Person Notified Date Time 3.8.2 Evaluate operating CA Pumps to cool CA System piping.

3.9 Ensure TURF released on the following:

CA Modulating Valves Reset Train A

CA Modulating Valves Reset Train B 3.10 Evaluate operability of CA System.

SRO End of Enclosure Unit I

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2010 MNS SRO NRC Examination QUESTION 79 2579 rk -

- KA_desc --_______

Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, SYSO76 procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 /43.5 /45/3 A2.O1 /45/13)ElLossofSWS Given the following plant conditions:

A loss of the Low Level Intake has occurred

AP-20 (Loss of RN) Case II (Loss of Low Level or RC Supply Crossover) has been implemented

RN is now aligned to the SNSWP

Train B of RN is now in service on both units

1) In accordance with AP-20, Train B RN pump flow is limited to less than

2) The RN pump flow is limited to prevent Which ONE (1) of the following completes the statement above?

A. 1. 11,500 GPM

2. RN pump impeller wear B. 1. 13,000 GPM

2. RN pump impellerwear C. 1. 11,500 GPM

2. vibration damage to the KC HX tubes D. 1. 13,000 GPM

2. vibration damage to the KC HX tubes Friday, May 14, 2010 Page 210 of 275

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 79 Jeneral Discussion __________________

In accordance with AP-20, Loss of RN Case II, Loss of Low Level or RC Crossover, the limit for Train A RN pump flow while aligned to the SNSWP is limited to less than 11,500 GPM and Train B RN pump flow is limited to less than 13, 000 GPM.

In accordance with the AP-20 Background Document, RN pump flow is limited to ensure NPSH is sufficient to prevent long-term RN pump impeller wear.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE: Part 1 is plausible because this is the flow restriction for Train A RN pumps while aligned to the SNSWP.

Part 2 is correct and is therefore plausible.

Answer B Discussion CORRECT. As explained above.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE: Part 1 is plausible because this is the flow restriction for Train A RN pumps while aligned to the SNSWP.

Part 2 is plausible because there are RN flow restrictions for the KC Hx to prevent vibration induced damage to the Hx tubes. However, this is not the basis for the flow restrictions on the RN pumps.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE: Part 1 is correct and therefore plausible.

art 2 is plausible because there are RN flow restrictions for the KC Hx to prevent vibration induced damage to the Hx tubes. However, this is not the basis for the flow restrictions on the RN pumps.

Basis for meeting the KA The K/A is matched because the unit has experienced a loss of all SWS as described in the stem of the question. The system has been placed in an alternate alignment as a result of this event (i.e. now taking suction from the SNSWP instead of the LLI).

Part B of the A (use procedures to correct control or mitigate) is met by the applicant having knowledge of the AP-20 restrictions on pump flow when aligned to the SNSWP. Part A of the KA (predict the impact) is met by the applicant having knowledge of the potential impact of exceeding the flow limits specified in AP-20. -____________

Basis for Hi Cog Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions (Rev I dated 03/11/2010) under the Screening Criteria for question linked to 10CFR55.43(b)(5) (Assessment and Selection of Procedures):

1) The question can NOT be answered by knowing systems knowledge. The limits for RN pump flow while aligned to the SNSWP are NOT specified in the RN System lesson plan and are NOT specified in the Operating Procedure (OP/I or 2/A164001006, Nuclear Service Water System). The only place these limits are specified is in AP-20. Additionally, the basis for the RN pump flow limits is NOT specified in the RN System Lesson Plan or the Operating Procedure. The basis for the flow restriction is only contained in the AP-20 Background Document.

Therefore, neither part of this question is systems level knowledge.

2) The question can NOT be answered by knowing immediate Operator actions.

3) The question can NOT be answered by knowing entry conditions for the AP.

4) The question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of AP-20.

5) The question requires the applicant to have knowledge of specific steps within AP-20 which are actions that are directed by the SRO during performance of the AP. -_______________________________________

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2010 MNS SRO NRC Examination QUESTION 79 Development Re fe Student References Provided -

)AP2OOO2 j.

References:

1) AP-20

2) AP-20 Background Document O76 Abi1i to () predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions,

___use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45/3 A2.01 /45/13)ElLossofSWS 401-9 Comments: arks!Status

]

Friday, May 14, 2010 Page 212 of 275

Question 79

References:

From AP-20:

rvNs LOSS OF RN PACE NC.

API 1N5500120 47 of 11 Crc II UNIT 1 Lssof Lo LeeI or 1C SLipply Crossover Re. 29 A::ION.. E:;:DIE: PESrONEE ;EE;CNEE ro

21. Maintain RN flow within operating limits as follows:

C:n Unit 1:

mp is runnnq.

trol;iit rtA RNbLrp fl

IF 6 RN pump is runninq. THEN ThROTTLE IRN-ISOD kNToD KC H> Ccntrol) to rnaintin 16 RN pump fox less :han 3000 GM.

C:n L nit 2

. if 24 RN pump is running, ThEN THROTTLE 2RN-894 (RN :c KC H Control.: to nuintain 24 RN pump fox Less than 11 SOD 3PM.

IF 26 RN pumo is runnino. THEN THRCTTLE 2RN-1EB RN To B KC I b Ccntrol) to niintain 2D N pump flox less than 13,000 .3D1.i NQIE LLI istheoni, availthle suctior for Rv punps.

22. IF RV pumps have lost suction, T-IEII ensure all RV pumps are in manual rnd off.

23. Enaure Contr& Roorri Prea Chiller in service PER Enc osure 4 (VCP(C Operationi.

From AP-20 Background Document:

AP)1 and 21A165001020 (Loss of RN)

CASE Ii STEP 21:

PuRPOSE:

Maintain the RN flow rate restrictions necessary to ensure NPSH requirements are met on the Pond suction source for the RN Pump (PIP 94-1429). Note short term operations a little above these NPSH flow rates is not an immediate concem. but rather a long term impeller wear issue DISCuSSION:

Refer to the below text from System Engineer John Pring from e-mail dated 3122/DO:

The numbers provided to you in the past by Jim are still accurate! but some clarification appears to be required. These flow limitations are based on engineering judgenient following evaluation of system testing (either flow balance data or pump perfoniance testing). The flow liniitauons are as follows:

Maximum Suggested RN Pump flow rate (any pump I any train) from Low Level Intake -- 16,000 gpm Maximum Suggested RN Pump flow rate (any A train pump) from SNSWP -- 11,500 gpm Maximum Suggested RN Pump flow rate (any B train pump) from SNSWP -- 13,000 gpm The term Suggested has real meaning in this context. f under emergency or abnormal conditions it becomes necessarg to achieve higher flow rates, then the operators should attempt to obtain the required flows. Suggested means that rnininial operating problems will be encountered if the flow thresholds are respected.

Please note that these flow rate limitations are pump related. Total Train flow rates will be significantly higher. You can use this note as documentation for the listed pump flow rate limitations John

REFERENCES:

PIP 94-1429 problem evaluation, remarks section & John Pring e-mail 3/22100 Page 35 of 40 Revs

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2010 MNS SRO NRC Examination QUESTION 80 25 KA KAdesc SYSO63 SYSO63 GENERICEIAbi1ity to perform specific system and integrated plant procedures during all modes of plant operati (CFR: 41.10/43.5/45.2/45.6) 2.1.23 With both Units at 100% RTP the following occurs:

Loss of Offsite Power occurred on Unit I

Both DGs started and loaded as designed

IA DIG subsequently trips on overspeed

At Step 17 of ES 0.1 (Reactor Trip Response), the decision is made to implement AP-07 (Loss of Electrical Power)

Which ONE (1) of the following describes the Time Critical actions directed by the CRS to mitigate this event per AP-07?

A. Complete Enc. 7 (DC Bus Alignment) to realign Battery Charger EVCA to Unit 2 within one hour.

B. Complete Enc. 7 (DC Bus Alignment) to realign Battery EVCA to Battery Charger EVCS within one hour.

C. Complete Generic Enc. 13 (VC and VA System Operation) to restart the Train A VCIYC Chiller within 37.5 minutes.

D. Complete Generic Enc. 13 (VC and VA System Operation) to swap Train A VCJYC Chiller power and water to Unit 2 and restart chiller within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 15 minutes.

Friday, May 14, 2010 Page 213 of 275

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2010 MNS SRO NRC Examination QUESTION 80 2580 3eneraI Discussion In the scenario described, Battery Charger EVCA would be off due to the tripped DIG IA. Step 20 directs the implementation of AP-07 Enc. 7 which provides direction for local actions to align EVCA Battery Charger to U-2 and restart the charger. This is required to be complete within one hour to prevent loss of the battery which is designed to provide power for one hour.

Answer A Discussion CORRECT: See explanation above Answer B Discussion INCORRECT: See explanation aboveBattery charger EVCS would only be utilized if it was aligned to the battery prior to the event. This would only be true if Battery Charger was out of service which is not indicted in the initial conditions.

PLAUSIBLE: Because it is the correct Enclosure, time frame and action if EVCS was in service. EVCS charger can be powered from either unit and be aligned to any vital battery so without familiarity with this enclosure this would seem a logical course of action.

Answer C Discussion INCORRECT: See explanation above. The Train A VCIYC Chiller is normally powered from U-l A Train Vital bus which is deenergized due to the failed 1A DIG.

PLAUSIBLE: Because it is the correct enclosure, correct action and Time requirement if the A VC/YC was available. This chiller can be powered from U-2 but is normally aligned to U-i. The time critical for manually selecting the other chiller if a VC chiller fails is 37.5 minuts Answer D Discussion INCORRECT: See explanation above. See explanation above.Train A VC/YC power and water only need to be swapped if a station blackout has occurred and that both D/Gs on one unit fail.

PLAUSIBLE: Because Step 17 of AP-17 directs implementation of this Enclosure within 30 mm and should the event be consistent with the need to perform this action, the enclosure and time requirements are correct. If a station blackout occurs and a VC chiller cannot be started due to a loss of power to the chiller, its power supply must be swapped to the opposite unit and the chiller started within i hour and 15 minutes from the time of the blackout.

asis for meeting the KA i(IA is matched the question is addressing actions required to recover from a reactor trip due to a LOOP associated with Unit 1 as directed by ES 0.1 and AP-7. In order to successfully answer the question the candidate is required to have a detailed integrated plant procedure knowledge associated with the DC electrical distribution system as well as knowledge of local Time Critical operator actions directed by these procedures along with the associated operational implications of not performing those actions within the given time constraints.

Basis for Hi Cog The analysis cog level is justified because the candidate must evaluate a given plant scenario, determine equipment availability and using procedural knowledge, determine the required course of action.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions (Rev 1 dated 03/11/2010) under the Screening Criteria for question linked to IOCFR55.43(b)(5) (Assessment and Selection of Procedures):

1) The question can NOT be answered by knowing systems knowledge alone.

2) The question can NOT be answered by knowing immediate Operator actions.

3) The question can NOT be answered by knowing AOP or EOP entry conditions.

4) The question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of the procedure.

5) The question requires the applicant to assess plant conditions and recall specific procedure content from AP-07. It requires the applicant to have an understanding of the specific procedural requirements associated with two different enclosures and the associated basis for those actions.

Therefore, this is an SRO level question.

Lb Level Cognitive Level QuestionType Question Source SRO Comprehension BANK MNS Bank NRC QIOO Development References Student References Provided P107, Loss of Electrical Power (Rev 28) page 9

.-07, Loss of Electrical Power Bkgd (Rev 7)

Pgsl4&l5 OP-MC-AP-07 Obj: 2 Friday, May 14, 2010 Page 214 of 275

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2010 MNS SRO NRC Examination QUESTION 80 258O KA KA_desc SYSO63 SYSO63 GENERICLIAbi1ity to perform specific system and integrated plant procedures during all modes of plant operation.

(CFR: 41.10/43.5/45.2/45.6) 2.1.23 401-9 Comments: RemarkslStatus Friday, May 14, 2010 Page 215 of 275

Question 80

References:

OP-MC-AP-07 Obj: 3 OBJECTIVES NNLLL OBJECTIVE L L P P 0 OORSR R00 Concerning AP101550017 (Loss of Electrical Power): X X X

. State the purpose of the AP

. Recognize the symptoms that would require implementation of the AP.

AP7001 2 Given scenarios describing accident events and plant x x x conditions, evaluate the basis for any caution, note, or step.

AP7002 3 State from memory the Immediate Action(s) and the x x x Response Not Obtained (RNO).

SCOO2

From AP-07 (Loss of Electrical Power) Pg 9 MNS LOSS OF ELECTRICAL POWER PAGE NO.

API i/A 550007 9 of 395 Cisc I UNIT 1 Loss of Normal Po:.er to Both I ETA and I ETS Ret. 28 A:::c:rE::ncTED aESPONSE RESrONSE 1121 CAUTIQN If operating train of VCIYC has failed, it may be time critical to swap operating VCIYC trains.

17. Initiate EPI1IN5000IG-1 (Generic Enclosures), Enclosure 13 (VC And VA System Operation) within 30 minutes of 810.

18. Check SJG Pressures STABLE OR

- IF AT ANYTIME SJG pressure goes down GOING UP. in an uncontrolled manner AND reactor tripped, THEN CLOSE the following valves:

. All MSIVs All MSIV bypass vales.

From AP-07 Bkgd Doc (Loss of Electrical Power) Pg 13 UNIT I CASE I STEP 17:

UNIT 2 CASE I STEP 15:

PURPOSE:

Ensure control room equipment temperature habitability is maintained.

DISCUSSION:

Excessive control room ambient temperatures could lead to redundant vital control room equipment failures. If the selected train chiller fails (or it power supply), the opposite chiller doesnt auto start. Its calculated in this type scenario if the opposite train chiller is started (must be done manually) within 37.5 minutes from the onset of the blackout, then equipment habitability will be maintained. If the enclosure is initiated within 30 minutes, then the few minutes it takes to diagnose the failure of the selected train, swap trains, and manually start the opposite train chiller can be accomplished in less than 7.5 minutes.

Other ventilation concerns (containment cooling, etc.) are addressed later in the procedure. This step was separated from them and moved earlier in the procedure (here) because of the potential time critical concern.

REFERENCES:

PIP M98-383 CaIc MCC 1211.00-33.0006

From AP-07 (Loss of Electrical Power)

MNS LOSS OF ELECTRICAL POWER PAGE NO.

API 1A550W07 iL of 395 Cse UNIT 1 Loss of Nomial Po.verto Both 1ETA and lETS Rev. 28 ACTICN/EXEOTED aEsc1isE F:s;Dr3E NDi 03Th:nE::

19. Check the following DC pumps start as required:

a. Check Unit 1 6900V busses AT ZERO

- a. Observe Caution prior to Step 20 and VOLTS. GQIQ Step 20.

b. Main Turbine EMERG BRG OIL b. Start pump.

PUMP.

c. DC S/U yAP EXTRACTOR. c. Start vapor extractor.

d. A CF PUMP TURB EBOP. d. Start pump.

e. B CF PUMP TURB EBOP. e. Start pump.

f. Check OAC AVAILABLE.

- f. Perfomi the following:

1) Dispatch operator to ensure Unit I AIR SIDE BACKUP pump running (Unit 1 Turbine Bldg. 760, 1 F-23).

2 Observe Caution prior to Step 20 and Q 12 Step 20.

g. Check computer point M 1D0581 (Ui

g. Dispatch operator to ensure Llnit 1 AIR Sen Air Side Seal Oil Backup Pump) - SIDE BACKUP pump running (Unit I ON. Turbine Bldg. 760. 1 F23I.

CAUTIQN If any battery charger has lost power; then restarting the charger(s) in Enclosure 7 (DC Bus Alignment) is time critical,

20. Have available licensed operator initiate Enclosure 7 (DC Bus Alignment) within 30 minutes of 5(0.

From AP-07 Bkgd Doc (Loss of Electrical Power) Pg 14 UNIT I CASE I STEP 19:

UNIT 2 CASE I STEP 17:

PURPOSE:

Ensure the major turbine building equipment (Main Turbine, CF Pumps, etc.) receive emergency DC powered lubrication following a loss of offsite AC power.

DISCUSSION:

This equipment may remain rotating for a period of time following the loss of offsite AC power. While rotating, lubrication is still required to prevent damage. Therefore, direction is provided to ensure the DC powered lubrication pumps are running to supply this lubrication. There is also a DC powered pump to ensure seal oil pressure to prevent a loss of generator hydrogen.

REFERENCES:

UNIT I CASE I STEP 20 CAUTION:

UNIT 2 CASE I STEP 18 CAUTION:

PURPOSE:

Cue the Operator the following step should have sufficient focus to ensure completion to meet the time critical nature of the step.

DISCUSSION:

The time critical nature of the step is: If power supply is lost to an essential battery charger (LOOP with failure of I DIG), it must be swapped to the other unit within an hour (MCC-1 381 .05-000-0220, 125VDC Vital Battery and Battery Charger Calculation).

The following step cues the operator to start the enclosure within 30 minutes. There is additional time critical assumptions made once the enclosure is initiated, so this caution ensures the operator is aware of these requirements.

REFERENCES:

Parent Question:

1 Pt Both Units are operating at 100% RTP.

Loss of Offsite Power occurred on Unit 1

Both DGs started and loaded as designed

1A DIG subsequently trips on overspeed

At Step 17 of ES 0.1 (Reactor Trip Response), the decision is made to implement AP-07 (Loss of Electrical Power)

Which ONE (1) of the following correctly describes the Time Critical local operator actions associated with AP-07?

A. Implement Enc. 7 (DC Bus Alignment) to realign Battery Charger EVCA to Unit 2 within one hour.

B. Implement Enc. 7 (DC Bus Alignment) to realign Battery EVCA to Battery Charger EVCS within one hour.

C. Implement Generic Enc. 13 (VC and VA System Operation) to restart the Train A VCIYC Chiller within 37.5 minutes.

D. Implement Generic Enc. 13 (VC and VA System Operation) to swap Train A VC/YC Chiller power and water to Unit 2 and restart chiller within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 15 minutes.

Proposed Answer: A

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2010 MNS SRO NRC Examination QUESTION 81 258l KA KA_desc SYSO 15 Ability to (a) predict the impacts of the following malfunctions or operations on the NIS; and (b based on those predictions, procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 / 45.3 A2.O1 I 45.5)LlPower supply loss or erratic operation Given the following conditions on Unit 1:

The unit is currently at 80% RTP with a power increase in progress

Power Range channel N-43 fails due to a faulty power supply

N-43 has been removed from service in accordance with AP-16 (Malfunction of Nuclear Instrumentation)

N-43 will be repaired in approximately 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

1. In accordance with Tech Spec 3.2.4 (QPTR), Quadrant Power Tilt Ratios shall be determined by

2. Quadrant Power Tilt limits prevent exceeding power distribution design limits.

Which ONE (1) of the following completes the statements above?

A. 1. calculation using the remaining three Power Range channels OR movable incore detectors

2. RADIAL B. 1. using the movable incore detectors ONLY

2. RADIAL C. 1. calculation using the remaining three Power Range channels QE movable incore detectors

2. AXIAL D. 1. using the movable incore detectors ONLY

2. AXIAL Friday, May 14, 2010 Page 216 of 275

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2010 MNS SRO NRC Examination QUESTION 81

.3eneral Discussion -__________________

With a power range channel out of service and power greater than 75% RTP, QPTR shall be determined by performing SR 3.2.4.2 (using incore detectors). If power was less than 75% RTP, QPTR is determined by calculation using the remaining three power range channels (SR 3.2.4.1).

However, surveillance SR 3.2.4.1 allows performance of SR 3.2.4.2. (using incore detectors) in place of 3.2.4.1.

In accordance with Tech Spec 3.2.4 Basis:

The QPTR limit ensures that the gross radial power distribution remains consistent with the design values used in the safety analyses.

Answer A Discussion INCORRECT. See explanation above.

PLAUSIBLE: Part 1 is plausible because this would be correct if power was less than 75% RTP.

Part 2 is correct.

Answer B Discussion CORRECT. See explanation above.

Answer C Discussion INCORRECT. See explanation above.

PLAUSIBLE: Part 1 is plausible because this would be correct if power was less than 75% RTP.

Part 2 is plausible if the applicant has the misconception that Quadrant Power Tilt limits prevent exceeding RADIAL distribution limits and AFD limits prevent exceeding AXIAL distribution limits.

Answer D Discussion nORRECT. See explanation above.

LAUSIBLE: Part 1 is correct.

Part 2 is plausible if the applicant has the misconception that Quadrant Power Tilt limits prevent exceeding RADIAL distribution limits and AFD limits prevent exceeding AXIAL distribution limits. -

Basis for meeting the KA The KA is matched because a power supply failure for an NIS channel has occurred and the applicant must determine the impact on Quadrant Power Tilt determination in accordance with Tech Spec requirements.

Basis for Hi Cog Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions (Rev 1 dated 03/11/2010) under the Screening Criteria for question linked to 10CFR55.43(b)(2) (Tech Specs):

1) It can NOT be answered solely by knowing < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs

2) It can NOT be answered solely by knowing the LCO/TRM information listed above-the-line

3) It can NOT be answered by knowing the Tech Spec Safety Limits or their bases

4) It requires the applicant to have specific knowledge of surviellance requirements (SR 3.2.4.1 & 3.2.4.3) which are below the line and are greater than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months surveillances. The applicant must also have knowledge of information contained in the Tech Spec 3.2.4 Basis Document.

Specifically, the reason for having Quadrant Power Tilt limits (i.e. to prevent exceeding RADIAL power distribution limits) is contained in the Tech Spec basis document and in the surveillance requirements.

Level Cognitive Level QuestionType Question Source SRO Memory NEW 49evelopment References Student References Provided earning Objective:

1) IC-ENB #19

References:

1) Technical Specification 3.2.4 QPTR Friday, May 14, 2010 Page 217 of 275

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2010 MNS SRO NRC Examination QUESTION 81 2581 42)_Technical Specification 3.2.4 Basis

. .A KA_desc SYSO 15 Ability to (a) predict the impacts of the following malfunctions or operations on the NIS; and (b based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5/43.5 /45.3 A2.O1 / 45.5)DPower supply loss or erratic operation 401-9 Comments: RemarkslStatus Friday, May 14, 2010 Page 218 of 275

Question 81

References:

From Tech Spec 3.2.4 QPTR:

QPTR 3.2A SURVEILLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1 ------

NOTES 1 With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER

<75% RTP, the remaining three power range channels can be used for calculating QPTR.

2. SR 3.2.4.2 may be performed in lieu of this Surveillance.

Verify QPTR is within limit by calculation. 7 days AND Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter with the QPTR alarm inoperable SR 3.2.4.2 ------

-NOTES Only required to be performed if input from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER75% RTP.

Verify QPTR is within limit using the movable moore 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months detectors.

McGuire Units 1 and 2 32.4-4 Amendment Nos. 184/166

From Tech Spec 3.2.4 Basis:

33QTR B 3.2.4 B 3.2 RWER DISTRIBUTION LIMITS B 3.2.1 QUADRANT POWER TILT RATIO itPTP)

BASES BACKGROUND The QPTR limil ensures thai the qross radial power distributon remains consistent hith the deign alues used in the safety analyses. Precise radial rower distthution measurements are iiade during startup testing, after rcfuclirg. and per odieclly diring power operation The power censit at arri pant in the cre must be limited so that the fuel design critea are nraintainod. Together, LCO 3.2.3. AXIAL FLUX DIFFERENCE iAFDl. LCO 3.2.4, and LCO 3.1.6 Control Rod Insertion Linits, proide limits on process ahabIes that ctaraclerize and control th three dinensinnl power diqtrdiiitioi of the rer.tor cnre Control of these ariaLdes ensures that the core operates wilhin the fuel desn criteria and that the power dstribution reniaiis within the bounds used in the safety analyses.

APPI l(ABI F This I CO prcltids core rinner rlistrihutionF thnt inbte the fnllning SAFETV ANALYSES fuel design criteria:

a. During a laige break loss of coolant accident (LOCA. there must be a hicjh level o probability that the peak dadding temperature does not eceed 22D0°F (Ret 1);

b. The DNBR calcLilated for the hottest fuel rod in the core mist be above the approved CNBR limit. (The LCO alone is nat suficiert to preclide DNB criteria violations for celain accidents, i.e., aDcidents in which the event itself changes the wre p>xer distributior. For these events, additional checks are made in the Dore reloac desijn process against the permissible statepoint power distrbLltiois.):

c. During an ejected rod accidsnt. the energy deposition to the fuel nust not eceed 28D cal/gm (Ref. 2): and

d. The control rods must be capable of siutting down the reactor wth a minimum requied Wl.l vdtli the highest worth contDl rod stuck 11Jlly y,ithdrami (Ret. ).

The LCO liniits on the ?.FD, the QPTR. the Heat flux I-lot Channel Fac:or (X,Y.Z)). the Nuclear Enthalpy Rise Hot Chanrel Factor (F(X.Y)),

0 (F

and control bank inserton are established to precude core power distributions that exceed the safety analyses limits.

McGuire Units I and 2 B 3.24-1 Revision No. 10

OPIR JL.

GAS CS ACTIONS cor1inuec reaching RP. As an added precaution, if the care poser does not reach RTP ithth 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . but is increased slowly, then the peaking fac:or surveillances must be performed wilhin 4E hours of the lime when the more restrictive of the pzser level limit de:ermined by Requied Action A. I o A.2 is exceded. These Conipletio Times are intenced to allow adequate tinie to nereaxe THERMAL POWER ta above the more restrictive limit of Required Action A.I or AZ while not perntting the core to rerain vith unconfirmed power distributions fr extended periods of trme.

Reuied A;tiurL A.? is iiiodiried In a Nule tItiiL sLites eiit the peuLiiiy factor surveillances must be done aler the exccre detectors have been calibrated to shc,v zero tilt (i.e., Required &ction A.Gi. The intent of this Note sto have the peaking factor surveillances perforiied at operating power levels, which can only be accomplished after the excore detectors are calibrated to show zero tilt and the core retu -ned to power.

6. I If Required Actiors Al through A..7are not completed within their associated Comp etion Times, the unit must be brought to a MODE or condition in which the requirements do not apply. To achieve this status, THERMAL must be reduced to 5O% RTP ithin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The allowed Completion Tine of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is r&sonable. based on operating experience regarding the amount of time required to reach the reduced power level without challengirg plant systems.

SURVEILLANCE SR 3.2.41 REQUIREMENTS R 3.2,4.1 is mo-c itied Lv to Notea Note I allows UPIR to be calcu ated with three power rnqe c iannels if THERMAL PCWER is c 75% RTP and the inputfrom one Po:ier range Neutron Flux charnel ia inoperable. Note 2 alloss performance of SR 32.4.2 in lieu of SR 3.2.4.1 if more than one iiput from Poser Range Neutron Flux charnels 2re inoperable.

This Surveillance vehfi that the OZTR. as indicated by the Nuclear lnstrument2tion System (NlSi excore channels, is within its limit& The Frcqucncy of 7 dave hon the QPT alarm is OPER\BLE is acccptablo because of the losv probability that this aIsn cart remain inoperable withojt detection.

McSuire Units I and 2 b 3.2.4b Revision No. IU

QPTR D

BASES SURVEILLANCE REQUIREMENTS (continued)

When the QPTR alarm is inoperable, the Frequency is increased to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This Frequency is adequate to detect any relatively slow chanqes in QPTR, because for those causes of OPT that occur quickly (e.g.. a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt.

SR 3.2.4.2 This Surveillance is modified b. a Note, which states that it is required ori[. when the input from one or more Power Range Neutron Flux channels are inoperable and the THERMAL POWER is 75% RIP.

With an NIS power ranqe channel inoperable, tilt monitorinq for a portion of the reactor core becomes dqraded. Larqe tilts are likely detected with the remaininq channels. but the capability for detection of small power tilts in some quadrants is decreased. Perfomiing SR 3.2.4.2 at a Frequency of (2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provides an accurate alternative means for ensuring that any tilt remarns within its limits.

For purposes of monitoring the QPTR when one power range channel is inoperable, the moveable incore detectors are used to confim that the normalized symmetric power distribution is consistent with the indicated QPTR and any previous data indicating a tilt. The incore detector monitoring is perfomed with a full incore flux map or Iwo sets of four thimble locations with quarter core symmetry. The two sets of four symmetric thimbles is a set of eight unique detector locations. These locations are C-B. E-5, E-11. H-3. H-fl, L-5, L-11. and N-B.

The symmetric thimble flux map can be used to generate symmetric thimble th. This can be compared to a reference symmetric thimble tilt, from the most recent full core flux map, to generate an incore tilt.

Therefore. incore tilt can be used to confirm that QPTP is within limits.

With one or more NIS channel inputs to QPTR inoperable, the indicated tilt may be changed from the value indicated with all four channels OPERABLE. To confirm that no change in tilt has actually occurred.

.,hich might cause the QPTR limit to be exceeded, the incore result may be compared against previous flux maps either using the symmetric thimbles as described above or a complete flux map. Nominally, quadrant tilt from the Surveillance should be within 2% of the tilt shown by the most recent flux map data.

McGuire Units 1 and 2 B 3.2.4-B Revision No. 10

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2010 MNS SRO NRC Examination QUESTION__82 KA KA_desc SYSO41 SYSO41 GENERICDKnow1edge of abnormal condition procedures. (CFR: 41.10 /43.5 /45.13)

Given the following conditions on Unit 1:

The unit is operating at 10-8 AMPS taking critical data

One condenser steam dump fails open

Crew is performing AP-Ol (Steam Leak)

Pressurizer level is stable

NC system temperature is 553° F and decreasing slowly

1. Based on the conditions above, to isolate the steam leak AP-Ol will direct the crew to

2. Isolating the steam leak is one of the design basis considerations for ensuring per Tech Spec 3.4.2, RCS Minimum Temperature for Criticality Basis.

Which ONE (1) of the following completes the statements above?

A. 1. trip the Reactor and close the MSIVs

2. the reactor remains subcritical in the event of a reactor trip B. 1. take A and B STEAM DUMP INTLK BYP switches to OFF/RESET

2. the reactor remains subcritical in the event of a reactor trip C. 1. trip the Reactor and close the MSIVs

2. proper indication and response of the excore detectors D. 1. take A and B STEAM DUMP INTLK BYP switches to OFF/RESET

2. proper indication and response of the excore detectors Friday, May 14, 2010 Page 219 of 275

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2010 MNS SRO NRC Examination QUESTION 82 2582

.,eneral Discussion the scenario given, a main condenser steam dump has failed open during a start up with the unit in Mode 2 holding power at 10-8 amps.

crew has implemented AP-Ol for a steam leak. Step 13 ofAP-Ol directs the crew to Check condenser dump valves CLOSED This would not be true so the RNO for this step directs the operators to select OFF RESET on Steam Dump Intk Bypass Channel A and B.

One of the major concerns of the CRS in this situation would be maintaining RCS temperature above the minimum temperature for criticality.

The question solicits this basis which includes the consideration that excore M would be adversely affected and if temperature were allowed to 1l_below this value, proper indication and required protective actions provided_would not be assured.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE: Part I is plausible because this would be the correct action if pressurizer level was decreasing (with maximum charging flow) or if NC system temperature was less than 55 1°F and decreasing.

Part 2 is plausible because it would be reasonable for the applicant to confuse the basis for Minimum temp for criticality with the basis for minimum temperature stated in many of our procedures which requires additional boration to prevent return to criticality.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE: Part 1 is correct and therefore plausible.

Part 2 is plausible because it would be reasonable for the applicant to confuse the basis for Minimum temp for criticality with the basis for minimum temperature stated in many of our procedures which requires additional boration to prevent return to criticality.

Answer C Discussion INCORRECT: See explanation above.

DLAUSIBLE: Part 1 is plausible because this would be the correct action if pressurizer level was decreasing (with maximum charging flow) or NC system temperature was less than 551°F and decreasing.

Part 2 is correct and therefore plausible.

Answer D Discussion CORRECT: See explanation above.

Basis for meeting the KA For this sencario, the applicant is given a malfunction of the Steam Dump System and is asked to demonstrate a knowledge of abnormal rocedure actions related to operation of the steam_dump controls to mitigate_the consequences_of the event. Therefore, the K/A is matched.

Basis for Hi Cog Eihis question is a higher cognitive level question because it requires the applicant to evaluate the plant conditions and determine the correct procedural actions based on the plant conditions.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures):

1) The question can NOT be answered solely by knowing systems knowledge. Either of these will stop the steam leak. However, plant conditions dictate the procedural flowpath requirements which will direct the crew to take the Steam Dump INTLK BYP switch to OFF/RESET versus tripping the Reactor and closing the MSIVs.

2) The question can NOT be answered by knowing immediate operator actions. Neither of the actions described are immediate actions.

3) The question can NOT be answered solely by knowing entry conditions for AOP or direct entry conditions for EOPs. These are detailed procedure steps from AP-Ol.

4) The question can NOT be answered solely by knowing the purpose, overal sequence of events, or overall mitigative strategy of the procedure.

This is detailed knowledge of procedure content related to knowing plant conditions that would require tripping the Reactor and closing the MSIVs as opposed to placing the steam dumps to OFF.

5) The question requires the applicant to analyze plant conditions and determine which section of the AP should be performed. Therefore, it is

[RO knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension BAiK BANK Q49l Friday, May 14, 2010 Page 220 of 275

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2010 MNS SRO NRC Examination QUESTION 82 2582 ioevelopment References Student References Provided -

(II

-01 (Rev 16) page (7 of42)

S 3.4.2 Basis KA KA_desc SYSO4 1 SYSO4 1 GENERIC Knowledge of abnormal condition procedures. (CFR: 41.10 I 43.5 I 45.13) 2.4.11 Comments: RemarkslStatus Friday, May 14, 2010 Page 221 of 275

Question 82

References:

From AP-Ol:

A;zIcn,.ExrzcTE: aE,3rcnsE ;EscNsz lt C. Operator Actions I. Monitor Foldout page.

2. Reduce turbine load to maintain the following:

. Excore Nrs LESS THAN OR EQUAL TO 1OU%

. NC Loop DTs LESS THAN OF DIE

. T-Avg-ATT-REF.

3. Check containment entry IN JQ Step 5, PROGRESS. -. ... *-.

1

4. Check steam leak KNOWN TO BE

- IF conditions warrant, THEN evacuate OUTSIDE CONTAINMENT. containment as follows:

a. An nounce All personnel evacuate Unit 1 containment.

Ix Actuate the containment evacuation al ami.

c. REFER TO RPIOIAIS700JO1 1 (Conducting a Site Assembly, Site Evacuation, or Containment Evacuation i as time alIos.

5. Check Pzr pressure prior to event - IF AT ANY TIME an 511 occurs due to GREATER THAN P-il (1955 PSIG). steam leak. THEN GO TO Enclosure 2 (511 Actions For Steam Break In Modes 3 and 4).

MNS STEAM LEAK PAGE NO.

AP! IIA5500?Ol $ of 42 UNIT 1 Re. 143 iIDLEXTEt .ESPCNSE ESCNSE OT C3TAI1ED

6. Check Pzr level: STABLE OR GOING UP.

a. icM1P flow less thai 00 GFIi at all tanes in subsequent

?ntia1 eJternate i f_:wpah.

b. I NV-238 Charing Liie FbI troh OPENING.

_c. PEN INV-241 Ui Seal .ater lnj low Control: ciIe aintning NC pump seal flo.; geter ha:i GGPM.

d. ce or -IJJL

e. NVpu f *eI qonc do.vii th aximurn rgrg f:ow, THEN ) IQ Step 9.

7. IF AT ANY TIME while in this procedure Pzr level cannot be maintained stable, THEN RETURNIQ Step 6.

_8. !QJ2Step12.

AcTION/EXPEcTED PZ3rDUEE ESPONSE NDT oaT:nr ci Perform the following:

NOTE If reactor trip breakers

. rzr j3ressiire p-ic-to - GREhTEj are closed and T-Avg is ThAN P-i I (3966 PS less than 553:F a feedwater isolation will occur in the next step.

a. Open reactor trip breakers.

b. CLOSE all MSlVs using individual valve F:tentia. a. te r nate pushbuttons.

fi:wpath.

c. If EP/i/Ai5000,ES-OJ Reactor Trip Response) has been implemented, Jiitti perform the following:

1) THROTTLE S/S feed flow to:

. coolJown

. Maintain at least one S?G N?R level greater than I 1%.

_2) GOTOStepii.

d. IF feednater isolation has occurred AND S/S levels going down in uncontrolled manner. ThiN perform the following:

1) Start CA pump(s).

2) WHEN desired to feed S(Ss wfth CM/CF. THEN REFER ID AP? liA/5500?D6 S/G Feedwater Nlalf Li nction

e. THROTTLE 5:5 feed flow to:

. Minimize cooldown

. Maintain at least one S/S N/R level greater than 11%.

_f. fiQIQStep 11.

rIIcu,r::;::T:: a:.s;:NSE ESCNE l:DT 3TAn:E:

10. and do Pctental aternate flcwp ath.

1 MSIVs ldIvlu,..

)fl S.

C.

ci. J1Recfl I. If Al MfY JlMf Rn level goes below 4%

AND cannot be restored using normal charging, Th[N perforni the following:

a. Ensure reactor tripped.

b. WHEN reactor tripped OR auto SI setpoint reached, THEN ensure S/I initiated.

c. Check Pzr pressure prior to e.ent -

c. GO TO Enclosure 2 (SI Actions For GREATER THAN P-Il (1955 PSlG Steam Break In Modes 3 and 4).

d. GO TO EP!1/N5000/E-0 (Reactor Trip or Safety Injectioni.

12. Announce occurrence on paging system.

PEcTED PE3PIUEE 7

kIIoN/E: AESrONSE NOT OBTAINED

13. Identify and isolate leak or Unit 1 as follows:

a. Check SM PORVs CLOSED.

- a. IF SIC pressure is less than 1092 P51G.

THEN perform the following:

1) CLOSE affected S/C SM PORV nianuaL loader.

2) IF SM PORV is still open, THEN perform the following:

a) CLOSE SM PORV isolation valve.

b) IF SM PORV isolation valve still open, TI-lEN dispatch operator to CLOSE SM PORV isolation valve.

b. Check condenser dump valves - Ix IF steam dumps required to be closed.

CLOSED. THUi perform the following:

1) SelectOFF RESET on the following switches:

STEAM DUMP INTLK BYPASS CHANNEL A

STEAM DUMP INTLK BYPASS CHANNELS.

2) IF valve will not close. THEN dispatch operator to CLOSE condenser dump valve isolation valve.

3) WHEN leaking condenser dump valve is isolated QR repaired. THEN return the following switches to

STEAM DUMP INTLK BYPASS CHANNELA

. STEAM DUMP INTLK BYPASS CHANNEL 5.

From TS Basis for 3.4.2 (RCS Minimum Temperature for Criticality):

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.2 RCS Minimum Temperature for Criticality BASES BACKGROUND This LCO is based upon meeting several major considerations before the reactor can be made critical and while the reactor is critical.

The applicant could misinterpret this passage to e first consideration is moderator temperature coefficient (MTC),

çç 3.1.3, Moderator Temperature Coefficient (MTC). In the transie imply a recriticallity concern nt analyses, the MTC is assumed to be in a range from sligh ie to negative and the operating temperature is assumed to the nominal operating envelope while the reactor is critical.

CO on minimum temperature for criticality helps ensure the plant ated consistent with these assumptions.

The second consideration is the protective instrumentation. Because certain protective instrumentation (e.g., excore neutron detectors) can be the correct answer affected by moderator temperature, a temperature value within the in the question. nominal operating envelope is chosen to ensure proper indication and response while the reactor is critical.

The third consideration is the pressurizer operating characteristics. The transient and accident analyses assume that the pressurizer is within its normal startup and operating range (i.e., saturated conditions and steam bubble present). It is also assumed that the RCS temperature is within its normal expected range for startup and power operation. Since the density of the water, and hence the response of the pressurizer to transients, depends upon the initial temperature of the moderator, a minimum value for moderator temperature within the nominal operating envelope is chosen.

The fourth consideration is that the reactor vessel is above its minimum nil ductility reference temperature when the reactor is critical.

APPLICABLE Although the RCS minimum temperature for criticality is not itself SAFETY ANALYSES an initial condition assumed in Design Basis Accidents (DBA5), the closely aligned temperature for hot zero power (HZP) is a process variable that is an initial condition of DBAs, such as the rod cluster control assembly (RCCA) withdrawal, RCCA ejection, and main steam line break accidents performed at zero power that either assumes the failure of, or presents a challenge to, the integrity of a fission product barrier.

APPLICABLE SAFETY ANALYSES (continued)

All low power safety analyses assume initial RCS loop temperatures the HZP temperature of 557°F (Ref. 1). The minimum temperature for criticality limitation provides a small band, 6°F, for critical operation below HZP. This band allows critical operation below HZP during plant startup and does not adversely affect any safety analyses since the MTC is not significantly affected by the small temperature difference between HZP and the minimum temperature for criticality.

The RCS minimum temperature for criticality satisfies Criterion 2 of 10 CFR 50.36 (Ref. 2).

LCO Compliance with the LCO ensures that the reactor will not be made or maintained critical (keff 1.0) at a temperature less than a small band below the HZP temperature, which is assumed in the safety analysis.

Failure to meet the requirements of this LCO may produce initial conditions inconsistent with the initial conditions assumed in the safety analysis.

APPLICABILITY In MODE 1 and MODE 2 with keff 1.0, LCO 3.4.2 is applicable since the reactor can only be critical (keff 1 .0) in these MODES.

The special test exception of LCO 3.1.8, PHYSICS TESTS Exceptions, permits PHYSICS TESTS to be performed at 5% RTP with RCS loop average temperatures slightly lower than normally allowed so that fundamental nuclear characteristics of the core can be verified. In order for nuclear characteristics to be accurately measured, it may be necessary to operate outside the normal restrictions of this LCO. For example, to measure the MTC at beginning of cycle, it is necessary to allow RCS loop average temperatures to fall below T 0 load, which may cause RCS loop average temperatures to fall below the temperature limit of this LCO.

Parent Question:

Initial conditions:

Unit 1 is operating at 10-8 amps taking critical data

One atmospheric steam dump fails opens

Crew is performing AP-Ol (Steam Leak)

What action is taken per AP-Ol to attempt to close the dump valve, and what design bases consideration (per Tech Spec 3.4.2, RCS Minimum Temperature for Criticality) is assured if this action is successful?

A. Take A and B STEAM DUMP INTLK BYP switches to OFF/RESET Steam generators are above their nil ductility reference temperature.

B. Dispatch operator to close the atmospheric dump valve isolation locally.

MTC will be in the range of slightly positive to negative.

C. Take A and B STEAM DUMP INTLK BYP switches to OFF/RESET The pressurizer is within its normal startup and operating range.

D. Dispatch operator to close the atmospheric dump valve isolation locally.

Proper indication and response of the excore detectors when the reactor is critical.

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2010 MNS SRO NRC Examination QUESTION 83 2583 KA KA_desc SYSOO2 Ability to (a) predict the impacts of the following malfunctions or operations on the RCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 /

A2.02 45.3 /45.5) Loss of coolant pressure Given the following conditions on Unit 1:

The unit is initially in MODE 3 with SD Banks withdrawn and NC System at full temperature and pressure

INC-32B (PZR PORV) fails open

AP-1 1 (Pressurizer Pressure Anomalies) has been implemented

The PORV isolation valve for 1 NC-32B will not close

An RO is directed to trip the Reactor and initiate Safety Injection

When attempted, both Reactor Trip breakers will not open Which ONE (1) of the following describes the impact of the conditions above on the requirements for initiation of Safety Injection AND subsequent procedural transition?

A. The crew shall wait until the Reactor is tripped and then initiate Safety Injection, even if the low pressure Safety Injection setpoint is exceeded.

The crew will then transition to AP-34 (Shutdown LOCA).

B. The crew shall wait until the Reactor is tripped and then initiate Safety Injection, even if the low pressure Safety Injection setpoint is exceeded.

The crew will then transition to E-O (Reactor Trip or Safety Injection).

C. The crew shall wait until the Reactor is tripped OR the low pressure Safety Injection setpoint is reached to initiate Safety Injection.

The crew will then transition to AP-34 (Shutdown LOCA).

D. The crew shall wait until the Reactor is tripped OR the low pressure Safety Injection setpoint is reached to initiate Safety Injection.

The crew will then transition to E-O (Reactor Trip or Safety Injection).

Friday, May 14, 2010 Page 222 of 275

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2010 MNS SRO NRC Examination QUESTION 83 L 2583 eneraI Discussion I In the scenario given with this question, the NCS has experienced a loss of pressure complicated by an ATWS (Failure of the RTBs to open from the C!R) Normally when approaching a ESF setpoint in a uncontrolled manor, the crew is expected to initiate the action prior to reaching the associated setpoint. In the case of an ATWS, the early initiation of SI would result in a FWI which could result in an extreme challenge to reactor safety (ATWS loss of Feedwater). In accordance with AP-1 1 Step 5 RNO, the crew should wait until the Reactor is tripped OR the SI setpoint is reached to ensure that SI is initiated.

The unit is in Mode 3 and the CLAs are not isolated so the correct transition is to go to E-0 (Reactor Trip or Safety Injection). If the unit was in Mode 3 with the CLAs isolated, the correct transition would be to AP-34 (Shutdown LOCA).

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE: Part 1 is plausible because initiating SI prior to the Reactor being tripped creates a worst case scenario ATWS event (i.e. ATWS with loss of feedwater). Therefore, it is reasonable for the applicant to conclude that SI should not be initiated until after the Reactor is tripped regardless of whether the SI setpoint is reached.

Part 2 is plausible because the unit is in Mode 3 and the appropriate transition would be to go to AP-34 if the CLAs were isolated.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE: Part 1 is plausible because initiating SI prior to the Reactor being tripped creates a worst case scenario ATWS event (i.e. ATWS with loss of feedwater). Therefore, it is reasonable for the applicant to conclude that SI should not be initiated until after the Reactor is tripped regardless of whether the SI setpoint is reached.

Part 2 is correct and therefore plausible.

Answer C Discussion riORRECT: See explanation above.

LAUSIBLE: Part 1 is correct and therefore plausible.

Part 2 is plausible because the unit is in Mode 3 arid the appropriate transition would be to go to AP-34 if the CLAs were isolated.

Answer D Discussion CORRECT: See explanation above.

Basis for meeting the KA The KA is matched because a loss of coolant pressure is occurring and the applicant must predict the impact of plant conditions on the procedural actions required to mitigate the event. The predicting the impact part of the KA is met because the applicant must determine how the procedural steps are different from the normal procedural flowpath based on a change in conditions (i.e. failure of the Reactor Trip breakers to open). The loss of pressure is responsible for the need to initiate SI.

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step. The applicant must first analyze the given conditions and determine that the unit is in a Mode 3 condition where the CLAs could not be isolated. The applicant must then recall from memory that the correct transition in Mode 3 with the CLAs NOT isolated would be to go to E-0 (Reactor Trip or Safety Injection) as opposed to AP-34 (Shutdown LOCA). The applicant must also evaluate the impact of a ATWS complicated by a loss of NCS pressure resulting in the need to

[initiate SI. This represents an analysis of the situation to determine that the normal expectations of a crew which approaching an ESF setpoint

[in an uncontrolled manor would not apply in the situation.

Basis for SRO only

[This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures):

1) This question can NOT be answered by knowing systems knowledge alone. This is strict procedure knowledge. This is not covered during systems training or discussed in a systems lesson plan.

2) This question can NOT be answered by knowing immediate operator actions. The immediate actions from AP- 11 address attempting to isolate the stuck open Pressurizer PORV. However, the actions to be taken by the crew as pressure continues to decrease are not part of the immediate actions.

3) This question can NOT be answered by knowing the entry conditions for AOPs. The steps to be taken by the crew are not based on the entry

.rnditions provided.

4) This question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of the AOPs. The question is based on knowledge of specific procedure content.

5) The question requires the applicant to have in-depth knowledge of specific steps wtihin AP- 11. Specifically, it requires the applicant to recall that the Step 5 RNO directs initiating SI only after the Reactor is tripped or the SI setpoint is reached. Additionally, the applicant must recall that Friday, May 14, 2010 Page 223 of 275

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2010 MNS SRO NRC Examination QUESTION 83 583 the same step directs transition to E-O if the unit is in Mode 3 or above with the CLAs not isolated. Therefore, this is SRO level knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided Learning Objectives:

1) N/A

References:

1) AP-l 1, Pressurizer Pressure Anomalies KA KA_desc SYSOO2 Ability to (a) predict the impacts of the following malfunctions or operations on the RCS; and (b) based on those predictions, procedures to correct, control, or mitigate the consequences of those malftinctions or operations: (CFR: 41.5 / 43.5 /

A2.02 45.3 / 45.5) ElLoss of coolant pressure 401-9 Comments: RemarkslStatus Friday, May 14, 2010 Page 224 of 275

Question 83

References:

From AP-Il (Pressurizer Pressure Anomalies):

FANS PRESSURIZER PRESSURE ANOMALIES PAGE N0 AP/! /4JS5D0? I I 2019 UNIT 1 Rev. JO a:::cn/Ez:cTEa s.roNsE ;is;QNSE 1ct OBT.a:Nz:

B. Synptoms

Pzr pressure channel failed

Pzr pressure going clown in an uncontrolled manner

Pzr pressure going up in an uncontrolled manner

Any Pu PORV or spray valve failed open

PZR PORV DISCS [IITEMP alarm

PRT SlUMP alarm.

C. Oierator Actions Check actual Pzr pressure I-lAS GONE Gi If) Step 17.

DOWN.

Check all Pr press tire channels -

jf either controlling channel is IN DICATING THE SAME. malfunctioning, J]iEtJ place PZP PRESS CNTRL. SELECT switch to backup channel.

Check PzrPORVs.CLOSED. Perform the following:

a, Cbss PORVs.

b. IF PORV .viIl not close. THEN close PORV isolation vaFve Check Pz. spray valves CLOSED.

- Perform the following:

a. Close Pn spray valve(s).

I

b. IF Y a reactortrip occurs AND spraj val.e still open. THEN stop 1A and 18 NC pumps.

-:T:DN:E:<n:To assP*CUSE LSOTE NCC O3TA::E:

5. Check Fzr PCRVs CLOSED,

- Perform the following:

a. II- assoaated PUttV isoIaton valve will riul clus pressuie yuiriy down rapidly. HEN:

1) Win Mcde 3 or above, prior to CL4 isola:ior. ThEN:

at Trip reactor b) WHEN reactor tipped DR auto SI etpoint reathet ThEN asure SI initialed.

c1 Eril.AsocoiE-a

Peactor Trip or Safety Injection).

Distracter PIausiil ity S..________________

b. Close asuciiied PCiRV iiiletdiiiii as follows:

. [F 1NC-326 (PZR PORVi failed, THEN close I NC-271 (PZR P:)RV Drn sal For 1 tsc-325).

. [F lNC-3iAPZR POPV) failed, THFN riasi 1 NC-77fl i:P7P PE)PI Drn sal For 1 NC-34A).

. IF I NG-33D i;.PZR PORV) tailed, fl EN close 1 NC-269 i:PZR PDRV Drn lsol For 1 NC-343B).

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2010 MNS SRO NRC Examination QUESTION 84 KA_desc PE0 15/017 [Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow):

-- (CFR 43.5/ 45.13)Abnormalities in RCP air vent flow paths and/or oil cooling system AA2.02 Given the following conditions on Unit 1:

The unit is operating at 100% RTP

1C NCP Oil Reservoir Level alarm is received on the OAC

Oil level indication on the OAC is -2.0 inches

IC NC pump motor bearing temperature is 200° F

AP-08 (Malfunction of NC Pump) Case II (NC Pump Motor or Motor Bearing Malfunction) has been implemented Which ONE (1) of the following describes the ACTIONS to be directed by the CRS in accordance with AP-08 and the HIGHEST POWER allowed at which the NCP can be stopped?

A. Trip the Reactor, verify reactor power less than 10%, then stop the 1C NCP.

B. Trip the Reactor, verify reactor power less than 5%, then stop the IC NCP.

(N C. Perform a unit shutdown to MODE 3 using AP-04 (Rapid Downpower), then stop the1CNCP.

D. Reduce the reactor power to < 48% using AP-04 (Rapid Downpower), then stop the 1C NCP.

Friday, May 14, 2010 Page 225 of 275

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2010 MNS SRO NRC Examination QUESTION 84 eneraI Discussion In accordance with AP-08, if an NC pump must be stopped reactor must be tripped if the unit is operating in Mode 1 or 2 and the NC can not be stopped until power is less than 5%.

If the NC pump trip criteria has not yet been exceeded but it is determined that the NCP still needs to be stopped, AP-08 directs performing a unit shutdown in accordance with OPI1/A16100/003 (Controlling Procedure for Unit Operation) or AP-04 (Rapid Downpower). The NCP is then stopped after all rods are inserted and the Reactor Trip breakers are open.

Answer A Discussion INCORRECT. See explanation above.

PLAUSIBLE: This answer is plausible because tripping the Reactor first is the correct action. Stopping the NCP when power is less than 10% is plausible since the NC low flow trips are defeated when less than 10% power (P-b).

Answer B Discussion CORRECT. See explanation above.

Answer C Discussion INCORRECT. See explanation above.

PLAUSIBLE: This answer is plausible because it would be the correct action if the applicant concluded that the NCP needed to be stopped but did not need to be stopped immediately.

Answer D Discussion INCORRECT. See explanation above.

PLAUSIBLE: This answer is plausible if the applicant concludes that the NCP needs to be stopped but does not need to be stopped immediately.

Reducing reactor power to less than 48% would allow the NCP to be stopped without causing a reactor trip (less than the single loop low flow trip setpoint). Then the unit shutdown to MODE 3 would be continued since power operation is not allowed without all NCPs in service.

Basis for meeting the KA he KA is matched because an malfunction of the oil cooling system has occurred and the applicant must determine the appropriate AP-08 actions based on plant conditions.

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step. The applicant must first analyze the given plant conditions and determine that the IC NCP must be stopped immediately. The applicant must then recall from memory the AP-08 actions required for stopping the NCP.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for screening questions linked to IOCFR55.43(b)(5) (Assessment and selection of procedures):

1) The question can NOT be answered by knowing systems knowledge. This is detail procedure content from AP-08.

2) The question can NOT be answered by knowing immediate Operator actions. None of the actions in the correct answer or in the distracters are immediate actions.

3) The question can NOT be answered by knowing entry conditions for the AP.

4) The question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of AP-08.

5) The question require the applicant to assess plant conditions and determine appropriate actions based on detailed knowledge of procedure content. Specifically, the applicant must determine that the NC pump needs to be stopped immediately which requires the Reactor to be tripped first and power verified less than 5% before the NCP can be stopped. Therefore, this is SRO level knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Memory NEW Development References Student References Provided Learning Objectives:

) N/A

References:

ri) AP-08, Malfunction of NC Pump Friday, May 14, 2010 Page 226 of 275

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2010 MNS SRO NRC Examination QUESTION 84 2584 KA KA_desc

( PE0 15/017 Ability to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow):

(CFR 43.5 /45.13)DAbnormalities in RCP air vent flow paths and/or oil cooling system LAA2.02 Comments: RemarkslStatus Friday, May 14, 2010 Page 227 of 275

4 Question 84

References:

From AP-08:

MNS MALFUNCTION OF NC PUMP PAGE NO.

AP/i/A5500/08 4of24 Ca II e

3 UNIT I NC Pump Motor or Motor Bearing Malfunction Rev. 12 ADTION EX?EDTED RE3ON.E RESEONS: NOT OBTAINET B. Symptoms

NC pump stator winding temperature going up

NC pump motor bearing temperatures going up

NC pump uppertlower oil reservoir level computer alarm.

C. OperatorActions

1. Check abnormal NC pump parameter - GO TO Enclosure I (Validation of NC KNOWN TO BE VALID. Pump Parameters).

2. Check NC pump parameters within If trip criteria valid, THEN Q T Step 5.

operating limits:

All NC pump stator winding temperatLires

- LESS tHAN 3ilF t

. All NC pump motor bearing temperatures

- LESS THAN 95°F

. All NC pump oil reservoir level computer points INDICATING BETWEEN

(-)1 .25 AND (÷)i .25.

3. IF AT ANY TIME any operating limit in Step 2 exceeded, ThiEt{QQTQStep 5.

_4. ]QStep5.

MNS MALFUNCTION OF NC PUMP PAGE NO.

API I /A/5506.OS 15 of 24 c.ase II UNIT 1 NC Pump Motor or Motor Bearing Malfunction Rev. 12 ACT ION/NESS DINE RESEONSE RESPONSE NOT OSTAINSO

5. Stop affected NC pump as follows:

a. [F A or S NC pump is the affected pump. THEN CLOSE associated spray valve:

. INC-27C (A NC Loop PZR Spray Control)

, 1NC-29C (B NC Loop PiP Spray Control).

b. Check unit status - IN MODE 1 OR 2. b. Perform the foflowing:

1) Stop the affected pump.

2) IF at NC pumps are oW THEM perform the foIlowhg:

a) Secure any boron dilution in progress.

b) jf in Mode 3, THEN immediately open Reactor Trip Breakers.

c) IF the step above results in rods dropping AND Pzr pressure is above P-I I. THEN GO TO I :A.5000/E-O (Pea ctor Trip or Safety Injection).

3) GO TO Step 6.

c. Trip reactor.

d. WHEN reactor power less than 5%,

THEN stop affected NC pump.

e. GQ TQ EP/11A15000!E-O (Reactor Trip or Safety Injection).

6. Announce occurrence on paging system.

MNS MALFUNCTION OF NC PUMP PAGE NO.

AP?liA55OD?0 Casell 17cf24 NC Pump Motor cc Motor Bearing Malfunction Rev. 12 UNIT I cTIONyE:PEcTEt RESFCNSE RESPONSE NOT csTa:N::

10. ffi$4fl Perfocm :hefoLlowing:

a.

Stator windirg temperalurea (UAC STABLE OR CCINC DOWJ

. ; temperatjs,. b.

ASLEIOR CcIINC D i Vibra:ion NDRMAL 11

. OP 1 IAIO []0?003 (Ccntrclling Picceduie Fur Unit Opeii[iun).

Fnctnssire$ 2 (Poser Perlrrrtinni Distracter OR Pausibility

2) WHEN in Mode 3. 4, or 5. THEN perfomi the following:

a IFaH JC pumps need to be stopped, RWN perform the to lowing:

1) Secure boron c)lutDn.

2 Do rot continue until rods inserted and reactor trp broakoro open.

b IF A cr B NC pump is the afected pun. JHEM CLOSE associated spra vaiwe:

. C A NC Loop PZR 7

INC-2 Spray Control

. 1NC-29C iS NC Loop PZR Spray Control:.

c) Stop affectec NC prp.

11. Check NC pumps ANY RUNNING.

- IF bolh ND pumps off AND 110 EP in effect Tll[N RLFER TO APIIIA/550(109 (Natural Circulatioli) as tine allows.

END

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2010 MNS SRO NRC Examination QUESTION 85 2585 KA KA_desc APEO22 APEO22 GENERICLIAbi1ity to verify that the alarms are consistent with the plant conditions. (CFR: 41.10 /43.5 / 45.3 /

45.12) 2.4.46 Unit I is operating at 100% RTP when the following alarms are received:

1AD-7 IJ1 (NC PUMP SEAL INJ LO FLOW)

1AD-7 /12 (REGEN HX LETDN HI TEMP)

1AD-7 I G2 (CHARGING LINE ABNORMAL FLOW)

The crew has implemented AP-12 (Loss of Letdown, Charging, or Seal Injection).

1. Based on plant conditions indicated by the alarms above, what actions are directed by AP-12?

2. What actions are directed by AP-12 regarding the restoration of letdown during the subsequent recovery?

A. 1. FIRST close the Letdown Orifice Isolations (1NV-458A, 457A, 35A) and then close 1 NV-lA, 2A (NC LID lsol To Regen Hx).

2. Pressurize the Letdown system locally.

B. 1. Close 1 NV-lA, 2A (NC LID Isol To Regen Hx) which will auto-close the Letdown Orifice Isolations (1 NV-458A, 457A, 35A).

2. Pressurize the Letdown system locally.

C. 1. FIRST close the Letdown Orifice Isolations (1NV-458A, 457A, 35A) and then close 1 NV-lA, 2A (NC L/D Isol To Regen Hx).

2. Pressurize the Letdown system from the Control Room.

D. 1. Close 1 NV-lA, 2A (NC LID Isol To Regen Hx) which will auto-close the Letdown Orifice Isolations (1NV-458A, 457A, 35A).

2. Pressurize the Letdown system from the Control Room.

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2010 MNS SRO NRC Examination QUESTION 85

..eneraI Discussion In accordance with AP-12, Loss of Letdown, Charging, or Seal Injection, the crew should first close the Letdown Orifice Isolations since they have indications that charging has been lost. Then because the Regen Fix Letdown Hi Temp alarm is in, they should close the isolations to the Regen Hx (NV-lA, 2A).

During subsequent recovery actions, the crew is procedurally directed to pressurize the letdown system from the control room since the Letdown Orifice Isolations were closed prior to NV-IA and 2A. Had NV-lA and 2A closed first, the crew would be required to pressurized the letdown system locally during restoration.

Answer A Discussion INCORRECT. See explanation above.

PLAUSIBLE: Part 1 is correct.

Part 2 is plausible if the applicant does not recall that the letdown line is pressurized locally when the Letdown Orfice Isolation Valves (1NV-458A, 457A, & 35A) close prior to 1NV-1A & 2A.

Answer B Discussion INCORRECT. See explanation above.

PLAUSIBLE: Part 1 is plausible if the applicant just remembers the step for closing 1NV-1A and 2A which states:

IF AT ANY TIME, REGEN HX LTDN HI TEMP alarm (IAD7-12) is LIT, THEN close the following valves:

1NV-IA 1NV-2A Procedurally the Letdown Orfice Isolations should have already been closed because the alarms in combination provide positive indication that a loss of charging has occurred and the steps to close the orifice isolations come before the steps to close NV-lA and 2A in the RNO column.

the applicant concludes that closing the Letdown Orifice Isolations first is the correct response then the Letdown Line would have to be

essurized locally in accordance with AP- 12 making Part 2 correct.

Answer C Discussion CORRECT. See explanation above.

Answer D Discussion INCORRECT. See explanation above.

PLAUSIBLE: Part 1 is plausible if the applicant just remembers the step for closing 1NV-IA and 2A which states:

IF AT ANY TIME, REGEN FiX LTDN HI TEMP alarm (1AD7-I2) is LIT, THEN close the following valves:

1NV-1A 1NV-2A Procedurally the Letdown Orfice Isolations should have already been closed because the alarms in combination provide positive indication that a loss of charging has occurred and the steps to close the orifice isolations come before the steps to close NV-lA and 2A in the RNO column.

Part 2 is plausible if the applicant does not recall that the letdown line is pressurized locally when the Letdown Orfice Isolation Valves (1NV-458A, 457A, & 35A) close prior to 1NV-1A & 2A.

Basis for meeting the KA The applicant must analyze the combination of alarms given in the stem of the question to determine the condition of the plant (i.e. in this case that a loss of charging has occurred). The applicant demonstrates that they have correctly identified plant conditions by selecting the correct actions from AP-12 for that plant condition. If the applicant choses the correct procedure actions based on their conclusions regarding plant conditions, they have demonstrated the ability to veri that the alarms are consitent with plant conditions. Therefore, the KA is matched.

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step. The applicant must first diagnose the conditions given to determine what has caused the alarms. The applicant must then recall from memory the procedure requirements for isolating letdown and the requirements for recovering letdown.

asis for SRO only nis question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures):

1) This question can NOT be answered by knowing systems knowledge alone. This requires tha applicant to analyze a given set of alarms and Friday, May 14, 2010 Page 229 of 275

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2010 MNS SRO NRC Examination QUESTION 85 2585 determine what plant conditions could have caused that combination of alarms. The applicant must then determine what procedural actions from

?-l2 are appropriate for plant conditions.

This question can NOT be answered by knowing immediate operator actions. AP-l2 has no immediate actions.

3) This question can NOT be answered by knowing the entry conditions for AOPs. The alarms given are entry conditions for AP-12 However, the applicant is given that AP-12 has been entered and determine what actions from the procedure are appropriate based on the combination of alarms.

4) This question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of the AOPs.

.5) This question DOES require the applicant to assess plant conditions (based on a combination of alarms) and determine from that assessment

the appropriate steps from AP-12 to be taken. This requires the applicant to have detailed knowledge of specific procedure steps from AP-l2.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided Learning Objectives:

1) AP 12002

References:

I) AP-12, Loss of Letdown, Charging, or Seal Injection KA KA_desc APEO22 APEO22 GENERICZAbility to verify that the alarms are consistent with the plant conditions. (CFR: 41.10/43.5 / 45.3 /

45.12) 2.4.46 01-9 Comments: Remarks!Status Friday, May 14, 2010 Page 230 of 275

Question 85

References:

From AP-12 (Loss of Letdown, Charging, or Seal Injection):

C. Op9ratorActicnc I. Check irdwIjIr IS dIItJiItitl LU IF kt cii vlIdIqiILq tl:icAlqli f,jeiieuLive Regenera:Ne lix follc*s: 1-Ix [as occurred. TI-tEN peiforrn the r blowing:

I Lharcinq how OFtAI hR I 1-IAN fl GPlil a. Ensu-e thee the fa[Ioing 9 akes are Cl CSFFi z 1NV_141 (I_I Seal Aater ltq Flos Contnl i THP.OaLED OPEN

- : NV-*1ESA i7E CPFA LID Orfice Outlct Cont lr,ol) 1N-244A ICharging Line Cctrit Outside Ccii) OPEN

- NV-457A (45 <3PM LID Di rice Outlet Cant Isol)

IN-2455 ICharging Lute Cant uLItside Isol) OPEN.

- : NV-3EA I2Qanable CD Ct-ifice Cuulel

/ *\ Ccnt lsal.L

/

h [P INV-7A (JO I ID lnl Tn Pcgin s While the applicant is not given closed, JFJE Q 10 Stcp 2.

nese specllc Irlalcatiors, they

c. I AT NY TIME ECEN HX LOON Ehotlid be abe to determine III TEMP alarrn( t,D-

, 2) It.

7 f-oni The cantintion oh arrns CLDS[ the tollo&ing vaCes:

hat charging flo:i has been lost.

F

NV-1A INC Lit sal In Regen Hx)

NV-2A (NC Lt sal Ta Regen Hx.

Letdown Recovery Actions from AP-12:

L2 Check the ft4kwiriy wikes - OPEN: Pci ILilil diti :

9 FoIloviri

_: 1N\-IA[NDJDl3oToReent) a. IF iurni1 letcown kno.n to be P 12 Step 49.

lN\-2& [ND JEt lsoi To Reqen -lx).

b. Hior to openng INV-IA Dr 1NV-2A in subsequent step. en sire that all personnel are oit lower cortannient of i:roteri ater hammer even:).

c. Obser.e Cau:io-i prior to Step t and GO ID Step 4$

_L3. GOTOStp42.

CAUTION Estatiishing wi-mel letdown without ocal pressurizttion nwy cause some water hammer.

I -

ARern ate 4 Di,wi mute ilcumiditiumis iiluw immmiuetlidw Path I eSLOtULIOtI of tiui inul letdowi i u fo1loiis:

_a. Check boh 1fW-1ANCL/DlsatTo [----)..a.

Reqen Hx) and 1NV-2A NC LO sot To Reqen Hx) -OPEN WITHIN TH[ LAST dli MINJ I tS.

h Check orilioe isoLation vake AUTD

- b. Peiloni the follaMnq:

CLOSED.

1 IF orth isdaon valves vere n,anuallv :lcsed while both INY- IA in-I NJ-A tre npn, T-IFN GO IOStejLS.

2 QjQRNforStep14A

MNS LOSS OF LETDOWN. CHARGING OR SEAL INJECTION PAGE NO.

APLIINSSDDI 12 21 of 43 UNIT 1 Rev. 22

44. Continued:i

c. Determine exact time each NV Ietdovm c. IF unable to determine time of closure.

val,e .cent closed on the OAC by IIIFN QIQ RNQ for Step 44.d.

performing the follozing:

1) Enter turn on code (ARCH VE 2 Ensure OAC automatically populates START TIME and STOP TIME. (previous hour:i.

3) Entergroup name AP1Z.

4) Click F3 VIEW P ID.

d. f orifice isolation valves reac d. Perform the following:

cosed PRIOR TO

-2 DLOINC 1) IF excess letdown is in service.

THEN obserce Caution and Note prior to Step 46 and GO TO Step 46.

NQIE Establishing normal letdown req uires local pressurization of letdown header.

Since this action takes significant time, establishing excess letdown first may be desired.

2) IF AT ANY TIME it is desired to establish excess Ietdo%vn, THEN GO JO Step 49.

3) Observe Caution and Note prior to Step 46 and GQIQStep $6, 45.

1NS LO3S OF LEDOVN, CH4RGING OR SEAL lNJEDTlOJ PAGE NO.

iP l55ODl2 22of4 TJMT 1 He*. 22 CAUTION It is pieferable to bcally essurize ihe letdown line pricr 10 estaNishin Ietcovm f1 due to possthle water Iiamniei.

NOTE If Dlantcorditions reqie rnnwditerestoiationof ncrnal letdown OSM may

,atve the requirement to loclf pres3urlze We tdo,n eader.

the app[ cant concludes that dosing the

40. IF ncrrnal Ietdjwn is iequ red prior to Onfice isolation Valves first is the correct Ic*collvpiestrizing Ietdcwn header, INO TO Step 18. response or does not recall we the letdovn Iie shDuld be pressurized ocaly relative tc the Dlcslre of the Letdown 47 I < sdtinns r1 th I tdo.ii Orifir sr.htinris) they wiiiI1 erwl lip tthL step a.

b.

C. _c. GOTOStep48.

d.

C I.

4Fi Fstahlish normal lctdnwn as follows: (j) Tfl tan

a. hnsure iNVb9(Ul Variable ULI is the beçinning of the sequence of Orthco Ottlct FIcw Cntr i i CLOSED. \ staps to pressiirie letdown from the Control Room. You end up here

b. Place I NV- 124 (Letdon Pressure regardless DI how long tdown was Cwrtrct in manual bemeen 1C-20% isolated provided INV-IA & 2A cosed OPFN

ç prior to the orilice isolatin vaIve.

c. Check the follow ng valves 0tN:

J. hnsure all rernreI are aut ot hr cntainrent prior to ecntinuing.

I Nv-i (hID LID bol To Reqen Hxi NV-2A tC LID bol To Regen F-lxl.

CAUTION A Pzr insurge wil occur when ch.nging flow is raised in next step.

letdown should he esuthlished withruadelayto limb 11w. aiiio,intnf insurge.

d. Eciabish cooling to ogcnerativc H by d. IF chrjrigflo to Rjoncrctivc Hx performing the fcllov,ing coicurrently: canrot be estaalisied. ThEN GO TO Step 49.

z Estublish ut leust 3E 3PM cliargirry flow by THROTL INC ODEW 1N23S (Charging Une Fla Control) or raising FL pump speed.

I ThP.O1TLE I NV-24 I Ul Seal Ver lnj Flo v Ccntroh b establish 9

approxiniately 5 (3PM seal injection flow to ach NC pump.

e. OPEN letdon line isolation vles as e. GO 0 Step 49.

fol lyws:

1) OPEN 1 NVTB Cont Outsice Isol).

2i OPEN 1NV-1A(NC L/E Isol To Rejen Hx).

31 OPFN I NV-2A (JC I IF knl To Regeri H>:).

4 OPEN 1kV-SEA (/criable LD Orifice Outlel Cait lso1

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2010 MNS SRO NRC Examination QUESTION 86 2586 KA KA_desc APEO27 Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions: (CFR: 43.5 I 45.13) L PZR heater energized/de-energized condition AA2.1O____________________________________________

Given the following conditions on Unit 1:

1A and 1D Pressurizer heater group supply breakers open at 1100 on June 1 due to a lightning strike and cannot be reclosed

A reactor startup is in progress with reactor power at 1% RTP

Heater groups 1 B, and I C are available Which ONE (1) of the following describes the required actions per Tech Spec 3.4.9, (Pressurizer)?

REFERENCE PROVIDED A. Restore PZR heater group 1A ONLY to operable status.

B. Restore PZR heater group 1 D ONLY to operable status.

C. Restore PZR heater group IA AND 1D to operable status.

D. Be in MODE 4 in 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months in accordance with Tech Spec 3.0.3.

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2010 MNS SRO NRC Examination QUESTION 86 L586 General Discussion The 1A and 1D Pressurizer heater groups have been de-energized. The remaining two groups of heaters are capable of maintianing Pressure under normal conditions. However, the OPERABILITY requirement is based on two groups of heaters with a capacity of 150KW each with each group being capable of being supplied by off-site or emergency power. Since Pressurizer heater groups IA and lB are the only groups which can be supplied by emergency power, both groups are required to be OPERABLE in Modes 1, 2, and 3. Therefore, PZR heater group 1A must be returned to operable status. Group 1D is not required for TS operability.

Answer A Discussion CORRECT. See explanation above.

Answer B Discussion INCORRECT. See explanation above.

PLAUSIBLE: This answer is plausible if the applicant does not recall from the Tech Spec Basis which groups of heaters meet the operability requirements for Tech Specs. TS 3.4.9 requires two groups of heaters with a capacity of 150KW each with each group being capable of being supplied by off-site or emergency power. Since Pressurizer heater groups 1A and lB are the only groups which can be supplied by emergency power, both groups are required to be OPERABLE in Modes 1, 2, and 3.

Answer C Discussion INCORRECT. See explanation above.

PLAUSIBLE: This answer is plausible if the applicant does not recall from the Tech Spec Basis which groups of heaters meet the operability requirements for Tech Specs. TS 3.4.9 requires two groups of heaters with a capacity of 150KW each with each group being capable of being supplied by off-site or emergency power. Since Pressurizer heater groups 1A and lB are the only groups which can be supplied by emergency power, both groups are required to be OPERABLE in Modes 1, 2, and 3.

Answer D Discussion INCORRECT. See explanation above.

PLAUSIBLE: This answer is plausible if the applicant does not recall from the Tech Spec Basis which groups of heaters meet the operability quirements for Tech Specs. TS 3.4.9 requires two groups of heaters with a capacity of 150KW each with each group being capable of being supplied by off-site or emergency power. Since Pressurizer heater groups 1A and lB are the only groups which can be supplied by emergency power, both groups are required to be OPERABLE in Modes 1, 2, and 3. If the applicant concludes that 1A and 1D heaters are the two groups of heaters required for operability, the appropriate action would be to enter TS 3.0.3 since the is no action statement for two of the required groups being inoperable. For example, if Pressurizer heater groups 1A and lB were inoperable this would be a correct answer.

Basis for meeting the KA Strict knowledge of Pressurizer heater energized/de-energized conditions is RO level knowledge. However, given a condition were a group of Pressurizer heaters is de-energized and asking applicant to determine if the TS operability requirements for Pressurizer heaters are met and having them apply Tech Specs to determine an appropriate action raises the question to the SRO level while achieving a match to the interpret portion of the KA.

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step. It requires the applicant to recall from memory that the Tech Spec requirement for operable heaters only applies to those with emergency power supplies (i.e. Group 1A and 1B). The applicant must the correctly apply Tech Specs to determine the correct actions to be taken.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions (Rev 1 dated 03/11/2010) under the Screening Criteria for question linked to 10CFRS5.43(b)(2) (Tech Specs):

1) It can NOT be answered solely by knowing < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs. There are no actions in TS 3.4.9 which are required to be completed in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or less.

2) It can NOT be answered solely by knowing the LCO/TRIvI information listed above-the-line. This question requires the applicant to recall information from the Basic Document for TS 3.4.9 to correctly apply the specification.

3) It can NOT be answered by knowing the Tech Spec Safety Limits or their bases. All actions are associated with TS 3.4.9, Pressurizer.

4) It DOES require the applicant to apply required actions and have additional knowledge contained in the Tech Spec Basis (specifically what constitues Pressurizer heater operability) to be able to apply the specification correctly and arrive at the correct answer.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Friday, May 14, 2010 Page 232 of 275

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2010 MNS SRO NRC Examination QUESTION 86 2586 Development References Student References Provided arning Objectives: Tech Spec 3.4.9, Pressurizer

) PS-NC#24

References:

1) Tech Spec 3.4.9, Pressurizer

2) Tech Spec 3.4.9 Basis KA KA_desc APEO27 Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions: (CFR: 43.5 I

45. 13)EIPZR heater energized/de-energized condition AA2.1O 401-9 Comments: RemarkslStatus

[____

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Question 86

References:

From Tech Spec 3.4.9:

Pressurizer 3.4.9 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.9 Pressurizer LCO 3.4.9 The pressurizer shall be OPERABLE with:

a. Pressurizer water level < 92% (1600 if

); and 3

b. Two groups of pressurizer heaters OPERABLE with the capacity of each group > 150 kW.

APPLICABILITY: MODES 1, 2. and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Pressurizer water levei A. 1 Be in MODE 3 with reactor 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months not within limit, trip breakers open.

AND A.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B. One required group of B.i Restore required group of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months pressurizer heaters pressurizer heaters to inoperable. OPERABLE status.

C. Required Action and CI Be in MODE 3. 0 hours0 days 0 hours 0 weeks 0 months associated Completion Time of Condition B not AND met.

C.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months McGuire Units 1 and 2 3.4.9-1 Amendment Nas. 184! 166

From Tech Spec 3.4.9 Basis:

P ressLirizer B 3.4.0 BASES APPLICABLE In MODES 1, 2, and 3, the LCO requirement for pressurizer level to SAFETY ANALYSES remain within the required range is consistent with the accident analyses.

Safety analyses performed for lower MODES are not limiting. All analyses performed from a critical reactor condition assume the exLstence of a steam bubble and saturated conditions in the pressurizer. In making this assumption, the analyses neglect the small fraction of noncondensible gases normally present.

Safety analyses presented in the UFSAR (Ref. 1) do not take credit for pressurizer heater operation however, an initial condition assumption of the safety analyses is that the RCS is operating at normal pressure.

The maximum pressurizer water level limit satisfies Criterion 2 or io CFR 50.36 (Ref. 2). Although the heaters are not specifically used in accident analysis, the need to maintain subcooling in the long term during loss of offsite power, as indicated in NUREG-0731 (Ref. 3). is the reason for providing an LCO.

[CO The [CO req uiremont for the pressurizer to be OPERABLE with a water volume 1600 cubic feet. which is equivalent to 92%, ensures that a steam bubble exists. Limiting the LCO maximum operating water level preserves the steam space for pressure controL The [CO has been established to ensure the capability to establish and maintain pressure control for steady state operation and to minimize the consequences of potential overpressure transients. Requiring the presence of a steam bubble is also consistent with safety analysis analytical assumptions.

The LCO requires two groups of OPERABLE pressurizer heaters. each with a capacity ISO kW. capable of being powered from either the offsite power source or the emergency power supply. Only heater groups A and B are capable of being powered from the emergency power supply.

The minimum heater capacity required is sufficient to maintain the RCS near normal operating pressure when accounting for heat losses through the pressurizer insulation. By maintaining the pressure near the operating conditions. a wide margin to subcooling can be obtained in the loops. The amount needed to maintain pressure is dependent on the heat losses.

APP[ICABILITY The need for pressure control is most pertinent when core heat can cause the greatest effect on RCS temperature. resulting in the greatest effect on pressurizer level and RCS pressure control. Thus. applicability has been designated for MODES 1 and 2. The applicability is also provided for MODE 3. The purpose is to prevent solid water RCS McGuire Units 1 and 2 B 3.4.9-2 Revision No. 0

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2010 MNS SRO NRC Examination QUESTION 87 2587 KA KA_desc -_______________________________________________________________________

EPEO3 8 EPEO3 8 GENERIC D Knowledge of abnormal condition procedures. (CFR: 41.10 / 43.5 /45.13) 2.4.11 Given the following conditions on Unit 1:

NC system is in MODE 3

SIG tube leakage occurs on C SIG

AP-1 0 (NC System Leakage within the Capacity of Both NV Pumps), Case I (SIG Tube Leakage) has been implemented

The crew has reached the step in AP-1 0 to perform a rapid cooldown The target temperature for the cooldown is determined based on (1)

The basis for the cooldown to the selected target temperature is (2)

Which ONE (1) of the following completes the statements above?

A. 1. ruptured SG pressure

2. to ensure the NC system is 20°F subcooled following the NC system depressurization B. 1. lowest intact SG pressure

2. to ensure the NC system is 20°F subcooled following the NC system depressurization C. 1. ruptured SG pressure

2. to ensure NC system temperature is below the saturation temperature for the ruptured SG PORV lift pressure D. 1. lowest intact SG pressure

2. to ensure NC system temperature is below the saturation temperature for the ruptured SG PORV lift pressure Friday, May 14, 2010 Page 234 of 275

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2010 MNS SRO NRC Examination QUESTION 87 2587

-eneraI Discussion In accordance with the AP-lO Basis Document for Case 1 Step 25:

The principal goal of the AP is to minimize and eventually stop primary-to-secondary leakage. This step is designed to determine the target temperature that will establish sufficient subcooling in the NC so that the primary system will remain subcooled after NC pressure is decreased in subsequent steps to stop primary-to-secondary leakage.

Since, in order to stop this leakage, the NC pressure must be decreased to a value equal to the affected steam generator pressure, the temperature at which this cooldown is terminated is dependent upon the affected steam generator pressure. A table is constructed for various affected steam generator pressures showing the fluid temperature corresponding to 20°F subcooling at each of these pressures, including allowances for subcooling uncertainties. For consistency with the EPs, the target temperature should be based on the core exit TCs. The 20°F subcooling is provided as operating margin to accommodate fluctuations in NC temperature, perturbations in affected steam generator pressure, interpolation between listed affected steam generator pressures, and overshoot during NC depressurization. SIG pressure ranges were specified as human factors enhancement.

Answer A Discussion CORRECT. See explanation above.

Answer B Discussion NCORRECT: See explanation above.

PLAUSIBLE: Part 1 is plausible if the applicant does not recall which pressure is used to determine the target temperature as they might conclude that the ruptured SG would be at a higher pressure (due to the NC system inleakage) than the intact SGs and that it would be appropriate to chose the intact SG with the lowest pressure to determine the target temperature for the cooldown. This would yield the lowest target temperature and provide for the most subcooling.

Part 2 is correct.

Answer C Discussion JCORRECT: See explanation above.

PLAUSIBLE: Part 1 is correct.

Part 2 is plausible because part of the strategy in AP-lO with regards to isolating the ruptured SG and conducting the cooldown is minimizing the possiblity of lifting the SG PORV on the ruptured SG. For example, during the isolation of the ruptured 5G. the basis for closing the MSIV last is to minimize the time between when the SG is bottled up and commencement of the rapid cooldown to minimize the possibility of lifting the PORV. So, it is plausible for the applicant to conclude that the target temperature is selected based on establishing conditions which will not result in the PORV lifting.

Answer D Discussion INCORRECT: See explanation above.

PLAUSIBLE: Part 1 is plausible if the applicant does not recall which pressure is used to determine the target temperature as they might conclude that the ruptured SG would be at a higher pressure (due to the NC system inleakage) than the intact SGs and that it would be appropriate to chose the intact SG with the lowest pressure to determine the target temperature for the cooldown. This would yield the lowest target temperature and provide for the most subcooling.

Part 2 is plausible because part of the strategy in AP-lO with regards to isolating the ruptured SG and conducting the cooldown is minimizing the possiblity of lifting the SG PORV on the ruptured SG. For example, during the isolation of the ruptured 5G. the basis for closing the MSIV last is to minimize the time between when the SG is bottled up and commencement of the rapid coo ldown to minimize the possibility of lifting the PORV. So, it is plausible for the applicant to conclude that the target temperature is selected based on establishing conditions which will not result in the PORV lifting.

Basis for meeting the KA The K/A is matched because the applicant must have detailed knowledge of steps from the abnormal procedure for dealing with SG Tube Leaks (AP-lO, NC System Leakage Within the Capacity of Both NV Pumps Case 1, Steam Generator Tube Leakage) and knowledge of the basis for steps from the AP.

Basis for Hi Cog asis for SRO only fhis question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev I dated 03/11/2010 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures):

1) The question can NOT be answered by knowing systems knowledge. This is knowledge of detailed procedure content.

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2010 MNS SRO NRC Examination QUESTION 87 2587

2) The question can NOT be answered by knowing immediate Operator actions. There are no immediate actions in AP-10.

The question can NOT be answered by knowing entry conditions for the AP- 10.

The question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of AP- 10.

5) The question requires the applicant to have knowledge of detailed procedure content from AP- 10 (specifically what parameter is used to determine the target temperature for the rapid cooldown) and the basis (from the AP- 10 Background Document) for performing the cooldown.

Therefore, this is SRO level knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Memory NEW Development References Student References Provided Learning Objective:

1) N/A

References:

1) AP-lO (NC System Leakage Within the Capacity of Both NV Pumps) Case I (Steam Generator Tube Leakage)

KA KA_desc EPEO38 EPEO38 GENERICEKnow1edge of abnormal condition procedures. (CFR: 41.10 / 43.5 / 45.13) 2.4.11 401-9 Comments: RemarkslStatus Friday, May 14, 2010 Page 236 of 275

Question 87

References:

From AP-lO:

NC SYSTEM LEAKAGE WITHIN THE O.P.AClT OF Pl1 _5]C 1D BCTHNPLMPE Steam Generator Tube Leakacje Re 22 UJT 1 C. Operator Actions

1. Check Pzr level - STABLE OR GOING UP. Perform the following as required to maintain level:

. rrt. chorclng fiow fln 200 GPF1 at alt times in sosepuem Qtepc.

h. E115ue 1 N1-23* iCharging Une Flo Conrol: OPENING.

c. OF-EN INV-24l :tJi Seal Water nj Flo Control) whLe maintohning NC pump seal flow reatertnan 6 GPM.

d. Reduce or isolate letdown.

e. Start additional N] pump.

t IF CLAs are isolated. AND P:r level is going down, TI-lEN peulorn the This proides ptau sibilitsi for following:

distracter 0. With the depressed PZR1e1elat5%,thiswouidseem _- 1) REFERjQEnclosure 1 ::Snudown a reaerabl ecurse but condsons Mode SGTL cticnc Tc Restore giwen in the stem do not support P:r Le9eI this path. 2: GO TO Step 2 g.. jf Pr leei cannot be maintained greater tan 4. OR P:r level going down with maximum charging flow.

THEN per1orn tne following:

1 IF 1 OR iC S)G klentilied at ruptrei THEN immediately d:o atdt io perto:o to oelat TO CA pump steam supply from ruptired 510 p Enclosure 3 (TO CA Pump Steam Sup pi solatic r.:.

2: Trip eactor.

3: WHEN reactortrpped OR auto SI setpoint reached. THEt-1 ensLre SI initiated.

_4: G.OTO EP1I.:A,SD00/ECiReactor Inn nr nf.r inrrhnr

From AP-lO NC SSTEFv1LEAKASESITHINTt-1ECAPACITh OF MIt . r FE 110.

6 of 129 Rev. 2_

C%IT Steant SeitratorTuLte Leakage

czz::zz:zz:::zctzz:::

7. Checl if Linit shutdown or reactGr trip requi.ed as fotlows:

a. Check .CT makeup -N PROGRESS. a. GOTO Step 7.c.

b. Check OT level- GOING UP. b. Assume St tube leak ccc in ne.t step is qreater than 9D GPM.

c. Check 515 tibe leak sze LESS THAN c Perform the foLowiric:

9Q FrI.

Kigr; N: punips tc FPST as ai OPEN the fcUowng nhea:

1 PunlpaSjct

1N:-221A;N From FVvST:

. IN.*-2225 NV Pumps Suet Foni rA ST.

b CLOSEthefcllowing alves:

1ti,14! CT uutlet IscI

!H:-4D5 :.:CT Outlet sd:,

2: IF reactc tlD hreaers c ose Tl-IEFJ perform the ¶cllcwing:

a Stat coth MD CA p-u nips.

7his is the correct answer C. Direction is given here to go to Step 0 tiich will b i imp reactor.

keeplliecrewinAp-lOforthe recovery r--- c IF Pr pressure was less tnan 19E PSIGcortothicer.t.

THEN 3Q TO Step .

This direction provides plausibitty for dstracter . H pressure was above P-li 1055 psig). mis wotld be the correct answer.

j IF reactc. trft bre,k era are open.

THEN GO TO Step 9.

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 88 25 KA KA_desc APEO58 Ability to determine and interpret the following as they apply to the Loss of DC Power: (CFR: 43.5 / 45.13) 125V dc bus voltage, low/critical low, alarm L_____________________________________________

Given the following conditions on Unit 1:

The unit was operating at 100% RTP when a total loss of onsite and oftsite power occurred

1. In accordance with AP-1 5 (Loss of Vital or Aux Control Power), what is the MINIMUM voltage on the DC Vital busses which requires the Vital Batteries (EVCA, EVCB, EVCC, EVCD) to be removed from service?

2. After power is restored and the battery chargers are placed in service, in accordance with Tech Spec 3.8.4 (DC Sources Operating), what is the MINIMUM voltage required for the Vital Batteries to be OPERABLE while on float charge?

A. 1. IlOvolts

2. l2Svolts B. 1. lOSvoIts

2. l25volts C. 1. ilOvolts

2. ilOvolts D. 1. lO5volts

2. liOvolts Friday, May 14, 2010 Page 237 of 275

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2010 MNS SRO NRC Examination QUESTION 88 L 2588

.eneraI Discussion In accordance with Tech Spec 3.8.4 Basis (DC Sources Operating):

The minimum battery terminal voltage limit is greater than or equal to 125 V while on float charge as discussed in the UFSAR, Chapter 8 (Ref.

4).

In accordance with AP-15 (Loss of Vital or Aux Control Power) the Battery EVCA Switch must be opened if Bus EVDA voltage decreases to 105 volts.

Answer A Discussion 1j6ORRECT. See explanation above.

PLAUSIBLE: Part 1 is plausible because because if the battery discharged to less than 110 volts surveillance SR 3.8.6.2 must be performed to verify that battery cell paramters are within limits.

Part 2 is correct.

Answer B Discussion RRECT. See explanation above.

Answer C Discussion 1i3ORRECT. See explanation above.

PLAUSIBLE: Both parts are plausible because if the battery discharged to less than 110 volts surveillance SR 3.8.6.2 must be performed to verify that battery cell paramters are within limits.

Answer D Discussion INCORRECT. See explanation above.

PLAUSIBLE: Part 1 is correct.

i/art 2 is plausible because if the battery discharged to less than 110 volts surveillance SR 3.8.6.2 must be performed to verify that battery cell paramters are within limits.

Basis for meeting the KA iiI(A is matched because the applicant must be familiar with the minimum (low/critical low) voltage at which the vital battery must be prated from the vital battery bus.

Basis for Hi Cog Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions (Rev 1 dated 03/11/2010) under the Screening Criteria for question linked to 10CFR55.43(b)(5) (Assessment and Selection of Procedures) and 10CFR55.43(b)(2) (Tech Specs):

Part 1:

1) The question can NOT be answered by knowing systems knowledge. The minumum voltage on the bus before the battery has to be removed from service is only addressed in AP-15. This voltage is NOT covered by the systems lesson plan or taught during systems training. Therefore, it is not systems knowledge.

2) The ques:ion can NOT be answered by knowing immediate Operator actions. There are no immediate actions associated with AP-15.

3) The ques:ion can NOT be answered by knowing entry conditions for the AP. The information tested does not constitue entry conditions for AP-15.

4) The question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of AP- 15.

5) The question requires the applicant to have knowledge of specific diagnositic steps within AP-l5. Specifically, if the minimum bus voltage is reached it requires the crew to transition to a section in the procedure to remove the battery from service.

Part 2:

11) It can NOT be answered solely by knowing < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs

,) It can NOT be answered solely by knowing the LCO/TRM information listed above-the-line

3) It can NOT be answered by knowing the Tech Spec Safety Limits or their bases

4) It requires the applicant to have knowledge contained in the Tech Spec Basis (specifically the minimum voltage limit for battery operability) wehe question correctly. --

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2010 MNS SRO NRC Examination QUESTION 88 258j Job Level Cognitive Level QuestionType Question Source SRO Memory BANK M1JS Exam Bank Question AP15NO1 Development References Student References Provided Learning Objectives:

1) AP15003

References:

1) Tech Spec 3.8.4 Basis (DC Sources Operating)

2) AP-15 (Loss of Vital or Aux Control Power)

KA KA_desc APEO58 Ability to determine and interpret the following as they apply to the Loss of DC Power: (CFR: 43.5 / 45.13)E1125V dc bus voltage, low/critical low, alarm AA2.02 401-9 Comments: RemarkslStatus Friday, May 14, 2010 Page 239 of 275

Question 88

References:

From Tech Spec 3.8.4 Basis:

t* turcesOperanIrg 5 3..4 CK2C ND ;conUnue:

Eaofl oattery:EVC& EVGb, ECC, EVCD1 has 3dequate siorace capacty t carry ihe requirsi duty cce ftr ane raur tsr the loss of the baiter charger ou:out q aOrllon the ati&ry is epaole of suppyiiq pce r for The operatior 01 artlpated niornen tary o during The one tdur perot.

Each I Vt battery Is separately housed n a jentlialed reom apart from Its charger and distrituton ceriers. Each channei is located Iran area separated phys cai and eieclrrcalyltvr the oir,ercharnei La ersure that a s igie iai ure ir cie subsystem does nor cause a aIiure in a relurdart subsyslern. There is no sharing bet.ceer reduniart Class IE subsystems such as catteries, battery chargers, or ilst.rDuIlon oarel&

Thc tzicr i TQr Th Nnrfl or D *re slzea to przuc reulr capacty at O% otnameptate ratng, canespondhg to ciarranted capacty atent oil[ie cy-cies ala tfle 10% de&gn derand Oattery size is based or 125% at required capacty ari, alter selection tf an awatiacle tcnrnral oattry. results ir a briery canacity Ir CXCS of 15Da or required capao ty. The in dividual cell oitage ml: is 2.1 3 V per *eii. I he nir ct1cllrtIt s reaier areuaIt 125 J e on rat o$iare as lisoussel n the U FAR, tnapier B iRef- s:.

The riter ibr iztig ar; ca ste rage tatie r cc are det1ed ir IEEE-4.95

.:Ret Ej.

Each oharirel er DC has arpe pcaeroutpui capacry fr1heslemiysbte operation ef connected icads required during normal operaflcri. w F e at Ihe sane Wile rnairt a nir.g ts battery bank 11.11 y charged. Each battery charger aso has sLtficiet capacty to restore tne battery fron the desigin nir rum rre ic Itelu y ciargei L1L witlir FQur 1illo ikzpiylr*g rormai steady state loads discussed ir the UFCAi9 Chao:er3 et 4.

APPLICABLE The II tat cone ttons or DesIgi Basis ACChd&9t DA ar transIent FEfl ANALYDES araiyses n the UFOAR, Chapler 6 :Ret 61, and ir the UFSA.9, Chantsr 15 [Ret ?, aaurns Ihat E gIeered Sflty Fealure OF sysie vs are 3PEft.5LE The CPERAS1L1fl cTtre DC sources is consistent wth the inital assumptions of the accient ana ys Es and $ based up n nieeiing the Cesin basis of the iniL This includes rairta nirg the DC s:uroes 2 PERABLE du ring CCC ien: ccriit ons in the eveit of:

MSGulrf UnIts 1 aM 2 0314-2 evIsIon No. 10

From AP-15 (Loss of Vital orAux Control Power):

MNS LOSS OF \qTAL OR AUX CONTROL POWER PAGE NO.

AP/ I /NSGDC 15 Cnclosure I Page 14 of 17 of 268 UNIT 1 Response To Dcgraded DC Bus Voltage e.

.z:zon.!zxrz:ra: :2:on&z rEnCflEE O OaTazNZD NOTE 1 n: distrbulion cente, .dtaçe goes down to ICE 107 Vots, the associat bat en.: ha niet Is dity cycle requirement. Further cepletion cf the hettery niav ri It in br,ttpr rhnng4 CpeninQ th nssn-Lited h,Sn, hmakr £hPn this oItage is reached ciII mpIeIelv denergize the associated distribution center.

17. IF AT AlLY liME dispatched opentor notifies Coiitrol Room that distribution center voltage reaches Iim.t in table below? THEM P rc indicated step to remoie batery from service:

DZ3TEUION VOLTAGE STE

. ,t1;+/-k LMII fl Stem 19 fl fl Step r.-:. 1:; 7:, Di EVDiE 1:5 SD 5721 1:3 7:Its Step 23 rIl IS V:lts Stet

18. Do net continue unkes dinctd y Step 11.

MNS LOSS OF VITAL OR AUX CONTROL POWER PAGE NO.

AP1A5500l5 65 of 268 Enclosure 1 -Page 16 of 17 UNIT I Response To Degraded DC Bus Voltage Rev. 20 TIQNEX9EDTED ES?ONE PE,OE OT CAINEE

21. IF AT ANY TIME EVDA Distribution Center reaches 105 volts, THEN evaluate performing the following:

a. Dispatch operator to open Distribution Center EVDA Compartment 2A (Battery EVCA Switch).

b. IF AT ANY TIME control power is needed to operate A Train breakers.

THEN contact station management to evaluate aligning battery to breaker control power circuits only.

c. Notify Unit 2 to perform the following:

1) WHEN EVDA deenergized, THEN REFER IAP2,A55OOi 5 (Loss of Vital or Aux Control Power) as time allows.

2) Trip Unit 2 reactor.

3) GOT0 EP/2N5000/E-O (Reactor Trip or Safety Injection).

d. WHEN EVDA deeriergized. THEN RETURN TO Step 1 in body of procedure as time allows.

e. Trip Unit 1 reactor.

f. GO TO EPIIA!50001E-O (Reactor Trip or Safety Injection).

22. IF AT ANYTIME EVOB Distribution Center reaches 105 volts. 11-lEN evaluate performing the following:

a. Dispatch operator to open Distribution Center EVDB Compartment 2A (Battery EVCB Switch).

b. Notify Unit 2 to GO TO AP2A-550015 (Loss of Vital or Aux Control Power).

c. RETURN T0 Step 1 in the body of the procedure.

Question 88 Parent Question:

Question 672 AP15NO1 AP15NO1 1 Pt Unit one was operating at 100% power when a total loss of onsite and offsite power occurred. Given the following events and conditions:

1 EVDA is supplying normal full power loads,

No battery charger is available,

Systems operate normally Which one of the following statements correctly describes the minimum length of time that bus I EVDA is designed to sustain loads and what action will protect the DC bus loads?

A. After 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the vital battery bus breaker will open automatically when bus voltage falls to 105 volts.

B. After 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the vital battery breaker must be manually opened when bus voltage falls to 105 volts.

C. After 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the vital battery breaker will open automatically when bus voltage falls to 107 volts.

D. After 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the vital battery breaker must be manually opened when bus voltage falls to 107 volts.

Answer 672 Answer: B Distracter Analysis:

A. Incorrect: the vital battery breaker does not automatically open Plausible: partially correct the design time for sustaining loads is 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months B. Correct: below this value the battery could be damaged or components will begin to fail.

C. Incorrect: the battery is expected to last for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and there is no automatic trip associated with low voltage Plausible: the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months requirement for battery performance is typical of the aux batteries voltage limit is 107 volts.

D. Incorrect: the vital batteries are not designed to sustain loads for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Plausible: partially correct DC bus protection is achieved by manually opening the breaker voltage limit is 107 volts.

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2010 MNS SRO NRC Examination QUESTION 89 2589 KA KA_desc APEO62 APEO62 GENERIC Li Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. (CFR: 41.10 I 43.5 / 45.12) 2.4.47 Unit 1 is operating at 100% RTP with B Train equipment in service when the following sequence of events occurs:

The Low Level Intake suction has been lost due to fouling associated with the intake grating

The crew is performing Enclosure 1 of AP-20 (Aligning B Train RN to Pond)

ORN-152 (Train lB & 2B Disch to SNSWP) failed to open and all attempts to move the valve have failed

The following SNSWP level trend is observed on the OAC:

741 739 .

737 735 733 731 SNSWP Level (ft) 729 727 725 723 721 4:00 5:00 6:00 7:00 8:00 9:00 10:00 11:00 Based on these conditions, the SNSWP level becomes INOPERABLE at (1)

The SNSWP minimum level ensures a sufficient volume of water to allow RN system operation for at least (2) following a design basis LOCA.

Which ONE (1) of the following completes the statements above?

A. 1. 0450

2. 5 Days B. 1. 1040

2. 5 Days C. 1. 0450

2. 30 Days D. 1. 1040

2. 30 Days Friday, May 14, 2010 Page 240 of 275

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2010 MNS SRO NRC Examination QUESTION 89 ieneral Discussion -

In the scenario the applicant is presented with a set of conditions where a loss LLI has resulted in the crew attempting to align the B Train of RN to the SNSWP. Due to the failure of a valve to move an abnormal alignment has resulted which is pumping water from the SNSWP to the lake and depleting the SNSWP inventory. Per TS 3.7.8 basis, the minimum required water level for the SNSWP is 739.5 which ensures a sufficient volume of water to allow NSWS to operate for 30 day following a design basis LOCA. With the SNSWP level trend provided in the stem, the applicant must differentiate between the TS level and the level at which SNSWP temperature is recorded (722).

Answer A Discussion INCORRECT: See explanation above PLAUSIBLE: Part 1 is correct and therefore plausible Part 2 is plausible because 5 days is the time the DIG is designed to operate with the minimum fuel required available. The applicant may misinterpret this to imply that the safe shutdown loads supplied would be required to be available for the same time frame.

Answer B Discussion INCORRECT: See explanation above PLAUSIBLE: Part 1 is plausible if the applicant recalls that the 722 elevation is discussed in the Basis Document forTS 3.7.8 but confuses this elevation as that which is required for minimum level.

Part 2 is plausible because 5 days is the time the DIG is designed to operate with the minimum fuel required available. The applicant may misinterpret this to imply that the safe shutdown loads supplied would be required to be available for the same time frame.

Answer C Discussion CORRECT: See explanation above.

Answer D Discussion INCORRECT: See explanation above LAUSIBLE: Part 1 is plausible if the applicant recalls that the 722 elevation is discussed in the Basis Document for TS 3.7.8 but confuses this 1evation as that which is required for minimum level.

Part 2 is plausible because 5 days is the time the DIG is designed to operate with the minimum fuel required available. The applicant may misinterpret this to imply that the safe shutdown loads supplied would be required to be available for the same time frame.

Basis for meeting the KA KA is matched because the candidate must evaluate the provided indications and diagnose the time at which the indicated SNSWP level is below that which is assumed in safety analysis. The utilizing the appropriate control room reference material in this case would be the use of the OAC level trend for SNSWP level.

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step. First the applicant must recall from memory the minimum level from the TS Basis assumed in the Safety Analysis. Then the applicant must analyze the SNSWP trend to determine at which time the SNSWP level decreases to less than the level recalled from memory.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03111/2010 for screening questions linked to 10CFR55.43(b)(2) (Tech Specs):

1) This question can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs. There are no actions that are 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or less associated with TS 3.7.8.

2) This question can NOT be answered by knowing information listed above-the-line. The information solicited by the question is contained in the surveillance requirements for the spec and is therefore not above-the-line information.

3) This question can NOT be answered by knowing the TS Safety Limits or their bases. The information tested is from TS 3.7.8, SNSWP

4) This question requires the applicant to recall information contained in the TS basis for SNSWP 3.7.8. He must determine if the given indication meet the minimum required level assumed in safety analysis and also recall the mission time for the SNSWP to provide a water supply for the NSWS.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Friday, May 14, 2010 Page 241 of 275

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2010 MNS SRO NRC Examination QUESTION 89 eIopment References Student References Provided 1 3.7.8 Basis AP-20 Enclosure 1 KA KA_desc APEO62 APEO62 GENERIC E Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate room reference material. (CFR: 41.10/43.5/ 45.12) 2.4.47 Comments: RemarkslStatus Friday, May 14, 2010 Page 242 of 275

Question 89

References:

From TS 3.7.8 Basis:

SNSWP 93.7:3

.4PPLICI\SLE SAFETY ANALYSES teonlinued:

cecay heat, and worst case single active failure {e.g., single laikxe cf a manrnace slruclure:i. The SNOWF is desi;necl inaccordarce with Regulatory Guice 1 27 Re:. 2). whch req uires a 30 ca SLipply of coolirg woter in the SNSWP.

Th 3NSWP srisfie (:ritednn 3 n 10 CFP 1 3i Rf 3)

LCO The 3NSP is required to be OPERABLE and is coraidered OPERABLE 1hi LII, btciiti II it c)ntains a sufficient vobme of water a: or below the naxirnurn rrtainin; the correct ..- temçeraturethat would aIlc: the NSVcS to opera for at least 36 days ansv,er .hich ;s 30 days. fulluwiriy [lie desiyri basis LOCA [liju1 Ilie lcssjI iet p.silive suUiuii

[earl (NPSI-1), and iithniit xreeditg Ihe triaximin, design telrerallJre Tnn bPPI ICkRI; 1W In MOflFS .3,3. anrl.t the5NSWPisrequirprltn.ciiprnrti OPERABILITY of tle equipnient serviced ovtne SNSWF Th :e.motconrrnnzthe Le UF-tRAELL in these MUL)LS. manmnmlnlwthan 339.

ft. The value foi the leve for the In MODE 5 orG the requirenieats of the SNSWParedeI recIuiredtenLPerSurel: 712 ft ntath psoJe p..su.aih 3 fin s sterns it sii-ts the applicau the

el÷tiu dasneter for aduquale.

AC1]ONS K1 If the SNSVvP is incperable the unit must be Flaced in a MODE ii vhich the LCD does not apple. To achieve this status, he unit nust be plccec in at least MODE 3 witin 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and in MODE S thin 36 [ours.

The allowed Completion Times are reasonable. based on operanq experierce, to reach the required unit :onditions fron full power cui idtiuns ii ai uiJei lvii iii ii ie arid wLI cut v ullerryn ig unit sys.enis.

SURVEILLANCE SR 3.7.3.1 REC UI REMENTS This SR veres that adequate long term (3D day i cooling can be

.,..#,-.,,4 Tk., ,-..-.&-,A L-,. ..-I ,. ,i-,.-. ... ,#,..* &ICQIJ

From AP-20 Enclosure I MNS LOSS OF RN RACE NO.

AP:IAJ55OOi2O Enclosure i-Page 1 of 4 UMT 1 Aligning B Train RN to Pond NOTh Shared RN valves can be operated from either units control switch.

Aiign B Train RN to SNSWP as follows:

a. Notify LI ii it 2 Operator that B Train RN will lSe aligned to the SNSWP.

Ii Check ORN-96 (Train lB & 2B SNSWP b. GQ TQ Case II (Loss of Low Level Supply OPER

- or RC Supply Crossover).

c. Open ORN- 1526 (Train 1 B & 26 Disc.h c. Perform the foLlowing to minimize pond to SNSWP). depletion:

1) Close the foLlowing va[wes:

. ORN-l lB (Train lB 8.26 Lii Supply)

ORN-IOAC Train IS & 26 LLI Supply;.

2) Dispatch operator to open ORN- 1526 (Train lB & 26 Disch to SNSWP) (aux bldg pipe chase, 716+2 EE-64, near Unit 2 containment surnp lines IS from cad door)

_3) WHffl4 QRN-1526 train iB&25 Disch to SNSVP) is open, TUN perform Steps 1 d through 1. h

4) Monitor SNSWP level as required to prevent reducinq SNSWP below Tech Spec leve[

(RNO continued on next page)

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2010 MNS SRO NRC Examination QUESTION 90 2590 KA_desc APE003 APEOO3 GENERICLIAbiIity to apply Technical Specifications for a system. (CFR: 41.10/43.2/43.5 / 45.3) 2.2.40 Given the following conditions on Unit 1:

Unit was at 100% RTP when rod M-4 dropped due to a blown fuse

AP-14 (Rod Control Malfunction) has been implemented

1) In accordance with Tech Spec 3.1.4 (Rod Group Alignment Limits), if the rod can NOT be restored to within alignment limits, power must be reduced to less than or equal to within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

2) Per AP-14 power must be reduced to less than a MAXIMUM of to retrieve the dropped rod.

Which ONE (1) of the following completes the statements above?

A. 1. 95%RTP

2. 75%RTP B. 1. 95%RTP

2. 50%RTP C. 1. 75%RTP

2. 50%RTP D. 1. 75%RTP

2. 75%RTP Friday, May 14, 2010 Page 243 of 275

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2010 MNS SRO NRC Examination QUESTION 90

,..eneral Discussion -______

In accordance with Tech Spec 3.1.4 (Rod Group Alignment Limits) the misaligned rod (in this case a dropped rod) must be restored to alignment limits with 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months OR SDM must be verified with limits AND power reduced to less than 75% RTP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

In addition surviellances for Enthalpy Rise and Heat Flux Hot Channel factors must be performed within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

In accordance with AP-14, power must be less than 50% RTP to retrieve the dropped rod. Additionally, AP-14 specifies power be reduced to less than 75% RTP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to comply with Tech Spec 3.1.4.

Answer A Discussion INCORRECT. See explanation above.

PLAUSIBLE: Part 1 is plausible because one of the actions required by Tech Spec 3.1.4 is to perform surveillance 3.2.2.1 for Enthalpy Rise Hot Channel factor determination. Normally this surveillance is required by Tech Spec 3.2.2 when power exceeds 95% RTP. Therefore it is plausible for the applicant to conclude that power needs to be reduced to less than this power so that the Enthalpy Rise Hot Channel factor surveillance can be perfonned.

Part 2 is plausible because power must be reduced to less than 75% RTP to comply with Tech Spec 3.1.4 if the rod can not be realigned within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Answer B Discussion INCORRECT. See explanation above.

PLAUSIBLE: Part 1 is plausible because one of the actions required by Tech Spec 3.1.4 is to perform surveillance 3.2.2.1 for Enthalpy Rise Hot Channel factor determination. Normally this surveillance is required by Tech Spec 3.2.2 when power exceeds 95% RTP. Therefore it is plausible for the applicant to conclude that power needs to be reduced to less than this power so that the Enthalpy Rise Hot Channel factor surveillance can be performed.

Part 2 is correct.

nswer C Discussion CORRECT. See explanation above.

Answer D Discussion INCORRECT. See explanation above.

PLAUSIBLE: Part 1 is correct.

Part 2 is plausible because power must be reduced to less than 75% RTP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to comply with TS 3.1.4. It is plausible for the applicant to conclude that the same restriction for reducing power due to the misaligned rod applies to realigning the rod as well.

Basis for meeting the KA Tech Spec 3.1.4 (Rod Group Alignment Limits) is the only TS that has applicability during a dropped rod scenario. Most of the actions contained in TS 3.1.4 are one hour or less actions making them RO level knowledge. One of the few actions that is not a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or less action requires the crew to reduce power to less than or equal to 75% RTP within 2 hrs if the misaligned rod (dropped rod in this case) can not be restored to within limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The applicant demonstrates the ability to apply Tech Spec 3.1.4 (Rod Group Alignment Limits) by recalling from memory the specific actions required to comply with the spec.

Basis for Hi Cog Basis for SRO only Part 1 of this question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev I dated 03/11/2010 for screening questions linked to 10CFR55.43(b)(2) (Tech Specs):

1) This question can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs.

2) This question can NOT be answered by knowing information listed above-the-line. This information is contained in the action statement section of TS 3.1.4.

3) This question can NOT be answered by knowing the TS Safety Limits or their bases. This question relates to TS 3.1.4 (Rod Group Alignment Limits)

) This question requires the applicant to have knowledge of actions required in the application of Tech Spec 3.1.4 (specifically the power duction required to comply with the spec)

Part 2 of this question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures):

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2010 MNS SRO NRC Examination QUESTION 90 2590 iie question can NOT be answered solely by knowing systems knowledge. The requirements for retreiving a dropped rod are not discussed in limits and precautions or in the systems lesson plan. Therefore, this is NOT systems knowledge.

fhe question can NOT be answered by knowing immediate operator actions. The action to reduce power to less than 50% RTP to recover the dropped rod is not an immediate action.

3) The question can NOT be answered solely by knowing entry conditions for AOP or direct entry conditions for EOPs.

4) The question can NOT be answered solely by knowing the purpose, overal sequence of events, or overall mitigative strategy of the procedure.

This is detailed knowledge of procedure content.

5) The question requires the applicant to have detailed procedure step knowledge from AP-14 (specifically the power reduction required to retrieve the dropped rod). Therefore, it is SRO knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Memory NEW Development References Student References Provided Learning Objective:

1) AP-14 #4

References:

1) Lesson Plan OP-MC-IC-IRE KA KA_desc APEO03 APEOO3 GENERICLIAbi1ity to apply Technical Specifications for a system. (CFR: 41.10/ 43.2 /43.5 /45.3) 2.2.40 31-9 Comments: RemarkslStatus Friday, May 14, 2010 Page 245 of 275

Question 90

References:

From AP-14:

MNS ROD CONTROL MALFUNCTION PAGE NO.

AP,iI1N5500?14 EnclosLire I PageS of 23 UNIT 1 Response To Dropped or Misaligned Rod Ri3

12. Reduce reactor power below 50% prior to rod realignment as follows:

a. Checkon[one rod-MISALIGNED. a. cQIQStep 12.c.

b. Ensure reactor power is tess than 75%

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of rod rnisaliqnment to comply with Tech Spec 3. 1.4.

c. Reduce load as directed in subsequent steps until reactor power is less than 5(1% to compl with Reactor Engineering req u ireni ent&

i Observe the following limitations during poer red Li ct ion:

1) Do not move rods until IAE determines rod movement is available.

2: Borate as required during power reduct:on to maintain T-Avg at T-Ref.

3) Monitor AFD during load reducUon.

4 IF AT ANY TIME AFD reaches Tech Spec limit 1Q reactor power is greater than 50%, IH,EN perform the following:

a) Trip Reactor.

bi GO TO EPLI A:5ooo:Eo (Reactor Trip or Safety Injection I

e. Reduce reactor power to less than 50%

RR one of the following procedures:

OP?iLA6 lOWDO3 lConttlling Procedure For Unit Operation).

Enclosure 4.2 (Power Reductioni OR AR: 1,?A:55Xr04 (Rapid Downpowerl.

From TS 3.1.4 Basis:

Rod Group Alignment Limits B 3.1.4 BASES ACTIONS (ntinued)

In many cases, realigning the remainder of the group to the misaligned rod may not be desirable. For example. realigning control bank B to a rod that is misaligned 15 steps from the top of the core would require a significant povver redLiction, since control bank D must be moved fully in and control bank C must be moved in to approximately lDQ to 115 steps.

Power operation may continue with one RCCA trippable but misaligned, provided that 5DM is verified within I hour.

The Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months represents the time necessary for determining the actual unit 5DM and, if necessary, aligning and starting the necessar systems and components to initiate boration.

B.12. 613. 6.24, 8.2.5, and 6.2.6 For continued operation with a misaliqned rod. RTP must be reduced.

5DM must periodically be verified within limits, hot channel factors (F

2XX,Zi and F(X.Y must be venfied :;ithin limits, and the safety analyses must be reevaluated to confirm continued operation is permissible.

Reduction of power to 76% RTP ensures that local LHR increases due to a misaliqned RCCA sill not cause the core .desiqn criteria to be exceeded Ref. 7 i. The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months gives the operator sufficient time to accomplish an orderly poser reduction sithout challenging the Reactor Protection System.

When a rod is known to be misaligned, there is a potential to impact the SUM. Since the core conditions can change with time, periodic verification of SDM is required. A Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient to ensure this requirement continues to be met.

Verifvinq that F {XX.Z and F(X.Y, are within the required limits 2

ensures that current operation at 75% RTP :dith a rod misaliqned is not resultinq in power distributions that may invalidate safety analysis assumptions at full power. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allo:s sufficient time to obtain flux maps of the core power distribution using the incore flux mapping system and to calculate FQ(X.Y,ZI and FNSH(X,Yi.

Once current conchtions have been verified acceptable. time is available to perform evaluations of accident analysis to determine that core limits will not be exceeded during a Design Basis Event for the duration of McSuire Units I and 2 B 1 1.4-6 Revision No. C

From TS 3.1.4:

RDd Groi. p A ignrnent Limits 3.1.4 ACTIONS (continued)

CONDITION PEQUIRED ACTION CDMPLETION4 TIME B. One rod not with n B. 1 Restore ic-cl to within 1 iour ciignnlert limits, a igninent linits.

OR B.2. .1 rif SDM 3 within the 1 iour limit specified in the OOLR.

OR B.2.:.2lnitia:e boration to restoc 1 10111 0DM to within limit.

AND B.2.2 Reduce THERMAL 2 iou is POWERto< 75% RTP.

AND B.2.3 Ve iP, 5DM is wilni I tIt Oitce pei limit specified in the DOLR. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND B.2.L Perform SR 12.1,1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND B.2.5 Perkiru SR 3.2.2.1. 72 liuuis AND B .2.6 Re-eqaltate safety S :lays analyses and confirn WSLI Its i eni.iirr valid br dLlralion of oDeratior under these coid itions.

(continued:I McGuire lJnits 1 and 2 3.1.4-2 Amendment Nos. 164/66

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2010 MNS SRO NRC Examination QUESTION 91 L<A APEO69 Ability to determine and interpret the following as they apply to the Loss of Containment Integrity: (CFR: 43.5 /

J45. 13) D Verification of automatic and manual means of restoring integrity AA2.02 Given the following conditions on Unit 1:

The unit is in MODE 5 following a refueling outage

PTIIIAI4200/002 C (Containment Closure I Integrity) is in effect

Both trains of ND are in service

Both ND pumps trip and cannot be restarted

AP-19 (Loss of ND or ND System Leakage) has been implemented Which ONE (1) of the following describes actions required byAP-19 based on the conditions above?

A. Notify the WCC SRO to dispatch Operators to isolate any open penetrations ONLY.

B. Evacuate Containment AND notify the WCC SRO to dispatch Operators to isolate any open penetrations.

C. Notify the Containment Closure Coordinator to initiate Containment closure ONLY.

D. Evacuate Containment AND notify the Containment Closure Coordinator to initiate Containment closure.

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2010 MNS SRO NRC Examination QUESTION 91 259l

.eneraI Discussion In accordance with AP- 19, if containment closure is in effect, AP- 19 will direct the Operators to evacuate containment, initiate a Site Assembly, and notify the Containment Closure Coordinator to initiate Containment closure.

Answer A Discussion INCORRECT. See explanation above.

PLAUSIBLE: It is plausible for the applicant to conclude that since the unit is in a refueling outage the WCC SRO has control over all work on containment penetrations and has the resources to get all open penetrations isolated. Since evacuating containment has nothing to do with containment isolation per se, it is plausible to conclude that evacuating containment is not required.

Answer B Discussion INCORRECT. See explanation above.

PLAUSIBLE: This is plausible because evacuating Containment is required. It is also plausible for the applicant to conclude that since the unit is in a refueling outage the WCC SRO has control over all work on containment penetrations and has the resources to get all open penetrations isolated.

Answer C Discussion 11/CORRECT. See explanation above.

PLAUSIBLE: This is a correct action however not complete actions. Since evacuating containment has nothing to do with containment isolation per se, it is plausible to conclude that evacuating containment is not required.

Answer D Discussion CORRECT. See explanation above.

Basis for meeting the KA The applicant is given a changing set of conditions which constitute a loss of containment integrity (because Containment integrity was not initially required and after conditions change it is required). The applicant is required to know how containment isolation is accomplished under is condition. Therefore, the KA is matched.

asis for Hi Cog Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev I dated 03/11/2010 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures):

1) The question can NOT be answered by knowing systems knowledge. How containment integrity is controlled during shutdown periods is not covered by any systems lesson plan. Therefore, this is not systems level knowledge.

2) The question can NOT be answered by knowing immediate Operator actions. There are no immediate actions in AP- 19.

3) The question can NOT be answered by knowing entry conditions for the AP. The actions for isolating containment in AP-19 are independent of the entry conditions.

4) The question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of AP- 19.

5) The question requires the applicant to have knowledge of detailed procedure content from AP- 19 (specifically the steps requiring Containment evacuation and containment closure). Therefore, this is SRO level knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Memory NEW Development References Student References Provided Learning Objective:

l)AP19002

References:

1)AP-l9, Loss of ND orND System Leakage Friday, May 14, 2010 Page 247 of 275

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2010 MNS SRO NRC Examination QUESTION 91 KA_desc APEO69 Ability to determine and interpret the following as they apply to the Loss of Containment Integrity: (CFR: 43.5 /

13) D Verification of automatic and manual means of restoring integrity AA2.02 401-9 Comments:

Friday, May 14, 2010 Page 248 of 275

Question 91

References:

From AP-19:

. Evaluate isolating *Dontoinn1nt as frillnws:

a Checc both ND purnp5 OFF - a. Perform tie followinq:

1) IF ak SZE greater than TO GP?L il-IhF% U) I(JStep&.

2) if leak oustd tip 2 3Iam snariy Coitin,ne,t or lJi,t veni EME.

I rHEN GO TO Step

) iFMANYIiMEInthNflpumps off. IH.E14 perfom, 3tep &b lhrouqh St

4) IL MAN! JIMt leak sIze greater lhrn D GFM OR leak caud trip 2 alami on am Drntainnieit or lint ient [fff EVEN perform Steps lb through Sf.

E tQTDStepG.

L 4\nnojnce the fdloicii-q on pae:

1 LiescrlptIcn ot event.

21 All persoinel evacuate Utit I conbi nment.

c. 4ctuste conftinrient vaelLqtizn alam..

d RFFFRTO PP/OL&wJrinmll lConductilxl a Ste Asse,tlv. She Evacuator. or Contahrrent Evacuator) while corrtinuirg Nith this p rocerlure.

e. Clit PT/l//d$200!002 C e. GOTOSLep6.

(Containment Closure) IN crrccT.

f. Notif CDn:ainnlenI Closuro (Doocliriator to nitate containnin1 closure.

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2010 MNS SRO NRC Examination QUESTION 92 2592 KAdesc

APEO76 APEO76 GENERICLIKnow1edge of abnormal condition procedures. (CFR: 41.10/43.5/45.13)

L24.1 Given the following conditions on Unit 1:

The unit is at 100% RTP

AP-18 (High Coolant Activity) has been entered due to 1EMF-18 (Reactor Coolant Filter 1A) in Trip 2 alarm Isotopic analysis of the NC system indicates the presence of Cobalt and Manganese which indicates that a (1) event has occurred and the required action in accordance with AP-18 to reduce the activity in the NC system is to (2)

Which ONE (1) of the following completes the statement above?

A. 1. failed fuel

2. place the Cation Bed demineralizer in service B. 1. failedfuel

2. increase letdown flow C. 1. crud burst

2. place the Cation Bed demineralizer in service D. 1. crud burst

2. increase letdown flow Friday, May 14, 2010 Page 249 of 275

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2010 MNS SRO NRC Examination QUESTION 92 ceneral Discussion From the basis document for AP-18:

Isotopes like Iodine and Cesium would indicate failed fuel, while isotopes like Cobalt and Manganese would indicate a crud burst.

For failed fuel events, one of the actions to reduce coolant activity is to place the Cation Bed Demineralizer in service. For a crud burst the appropriate action is to increase letdown flow.

Answer A Discussion

[NCORRECT. See explanation above.

PLAUSIBLE: This answer is plausible if the applicant is not familiar with the isotopes that would differentiate between a crud burst and failed fuel or the required actions from AP- 18 which are specific to a crud burst or failed fuel event.

Answer B Discussion ENCORRECT. See explanation above.

PLAUSIBLE: This answer is plausible if the applicant is not familiar with the isotopes that would differentiate between a crud burst and failed ue1 or the required actions from AP- 18 which are specific to a crud burst or failed fuel event.

Answer C Discussion INCORRECT. See explanation above.

PLAUSIBLE: This answer is plausible if the applicant is not familiar with the isotopes that would differentiate between a crud burst and failed fuel or the required actions from AP-l8 which are specific to a crud burst or failed fuel event.

Answer D Discussion CORRECT. See explanation above.

Basis for meeting the KA The KA is matched by Part 2 of the question in that the actions listed are from AP- 18 and are different for failed fuel as opposed to a crud burst.

art 1 of the question asks for information from the Background Document for AP-18.

Basis for Hi Cog This is a higher cognitive level question because it requires multiple mental steps. The applicant must first recall from memory (from the basis document) that the presence of Cobalt means that a crud burst has occurred. The applicant must then recall from memory the appropriate actions for a crud burst to reduce activity levels.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for screening questions linked to IOCFR55.43(b)(5) (Assessment and selection of procedures):

1) The question can NOT be answered by knowing systems knowledge alone.

Knowledge of the different isotopes which indicate failed fuel or a crud burst and the methods for reducing radiation levels associated with those events is not expected knowledge for ROs or SROs at MNS. Therefore, it is not systems level knowledge.

2) The question can NOT be answered by knowing immediate Operator actions. There are no immediate actions associated with AP- 18.

3) The question can NOT be answered by knowing AOP or EOP entry conditions. Knowing the entry conditions for AP-18 does not allow the applicant to answer this question.

4) The question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of the procedure.

5) The question requires the applicant to have detailed knowledge from the AP- 18 Basis document (specifically that the presence of Cobalt and Manganese indicate a crud burst) and detailed procedure content knowledge (i.e. the requirement to increase letdown as opposed to placing the Cation Bed demineralizer in service). Therefore, this is SRO level knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided earning Objectives:

1) AP18003

References:

1) AP-18

2) AP- 18 Background Document Friday, May 14, 2010 Page 250 of 275

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2010 MNS SRO NRC Examination QUESTION 92 (1 APEO76 desc APEO76 GENERIC D Knowledge of abnormal condition procedures. (CFR: 41.10 /43.5 / 45.13) 2.4.11 401-9 Comments: ksIStius Friday, May 14, 2010 Page 251 of 275

Question 92

References:

From AP-1 8 Background Document for procedure Step 3:

STEP 3 PURPOSE:

To ensure the mixed bed demineralizer thats normally in service is not depleted and to determine if the cause of the high activity is from a crud burst or from failed fuel.

DISCUSSION:

Step 3.a checks the decontamination factor (DF) of the mixed bed demineralizer. DF is the ratio of the Influent concentration divided by the Effluent concentration. The higher the DF, the more effective the demin for removing impurities. A DF of 100 is typical of a fresh mixed bed, and a DF of 10 or less is typical of a mixed bed near depletion. If the DF were low, it would be appropriate for Chemistry to request swapping to the standby mixed bed.

Step 3.b request Chemistry to run an isotopic analysis to determine cause of high activity.

Since they already have an influent sample in hand for determining DF, it can be used for this purpose. Isotopes like Iodine and Cesium would indicate failed fuel, while isotopes like Cobalt and Manganese would indicate a crud burst.

REFERENCES:

Primary Chemistry Lesson Plan OP-MC-CH-PC STEP 4:

PURPOSE:

Reduce redeposition of crud throughout the plant.

DISCUSSION:

At the normal letdown flow rate of 75 gpm, it takes almost 21 hours2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months to pass one entire volume of reactor coolant through the NV System. But a letdown flow of 120 gpm will circulate one entire volume of reactor coolant in approximately 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (at 120 gpm letdown flow, 50% of the crud is removed every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ).

REFERENCES:

Primary Chemistry Lesson Plan OP-MC-CH-PC

From AP-18:

MNS HIGh ACTIVITY IN REACTOR COOLANT PAGE NO.

APi 1/A15500.:l8 2 of 4 UNIT 1 Rev. 3 A DrICN/XE.TED RES?ONSE P.ES ON3A OT C:BIJE:r B. .$yirpms

1EMF.4 REACTOR COOLANT HI RAD alarm

IEMF-lS REACTOR COOLANT FILTER 1A alarm

1EMF-19 REACTOR COOLANT FILTER 1B alarm

Chemistry sample results indicate an unexpected increase in NC System activity.

C. Operator Actions

1. Place one Outside Air Pressure Filter train in service PER Enclosure I (Pressurizing the Control Room).

2. Check 1NV-127A (LiD Hx Outlet 3-Way Align valve to DEMIN position.

Temp Cntrl) ALIGNED TO DEMIN.

3. Determine cause of high activity as follows:

a. Request Chemistry to check decontamination factor of mixed bed dorn ineral izer.

b. Notify Chemistry to perform an NC System isotopic analysis to determine it high actiit is from a crud burst or tailed fuel.

4. IF AT ANY TIME itis determined that high activity is from crud burst, fl_EN raise letdown flow to 120 GPM PER 0P111A162001001 A (Chemical and Volume Control System Letdown).

Enclosure 4. (Establishing Maximum.

Normal Letdown).

PANS HIGH ACTIVITY IN REACTOR COOLANT PAGE NO.

AP/1 /A/5500/l 8 3 of 4 UNIT 1 Rev. 3 ADTICN: E:?EDTED P,ESTCNSE RISTONSE NOT CECAINED

5. EF AT ANYTIME it is determined that high activity is from failed fuel, THEN perform the following:

a. Ensure mixed bed demineralizer in service.

b. Notify Chemistry to consult with Reactor Group and RP to determine if the cation bed demineralizer should be placed in service.

c. 1EAI.ANYIIM.gChemistryrequests cation bed demineralizer be placed in service, THEN place in service PER ORl/A!6200!OQlD (Chemical and Volume Control System Dem ineralizers). Enclosure 4.3 (Removing/Returning the Cation Bed Demineralizer frorn:to Service).

d. REFER TO RP/O/A/5700/000 (Classification of Emergency).

e. Notify Reactor Group to discuss high activity in NC System with General Office Nuclear Engineering.

6. Notify Radwaste to ensure VCT H2 purge flow is established.

7. REF.EETQTech Spec 3.4.16 (RCS Specific Activity).

8. WHEN station management determines Control Room pressurization no longer required, THEN secure PER 0P101A16450101 I (Control Area Ventilation/Chilled Water System),

Enclosure 4.4 (Control Room Atmosphere Pressurization During Abnormal Conditions).

END

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2010 MNS SRO NRC Examination QUESTION 93 2593 KA_desc WEO3 WEO3 GENERICAbility to verify that the alarms are consistent with the plant conditions. (CFR: 41.10/43.5/45.3 /45.12) 2.4.46 Given the following conditions on Unit 1:

A Reactor Trip and Safety Injection have occurred due to a Small-Break LOCA inside Contaiment

Containment pressure peaked at 2.5 PSIG

ES-i .2 (Post LOCA Cooldown and Depressurization) has been implemented

Both ND pumps are running

NC system pressure is 250 PSIG and decreasing slowly Which ONE (1) of the following describes the FIRST FWST level and Containment Sump conditions that require stopping both ND pumps prior to swapping to the containment sump?

A. FWST level 200 inches Both CONT SUMP LEVEL GREATER THAN 2.5 FT alarms are DARK B. FWST level 260 inches Both CONT SUMP LEVEL GREATER THAN 2.5 Fl alarms are DARK C. FWST level 200 inches Both CONT SUMP LEVEL GREATER THAN 2.5 FT alarms are LIT D. FWST level 260 inches Both CONT SUMP LEVEL GREATER THAN 2.5 FT alarms are LIT Friday, May 14, 2010 Page 252 of 275

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2010 MNS SRO NRC Examination QUESTION 93 E2593 eneraI Discussion In accordance with ES-l.2, if FWST level decrease to less than 250 inches AND both CONT SUMP LEVEL GREATER THAN 2.5 FT alarms are DARK, and NS pumps are OFF, the ND pumps must be stopped prior to reaching 180 inches to prevent vortexing following suction transfer to the sump.

Answer A Discussion CORRECT. See explanation above.

Answer B Discussion INCORRECT. See explanation above.

PLAUSIBLE: The second part is correct.

First part is plausible because the level is less than the level at which FWST makeup is required.

Answer C Discussion See explanation above.

rNCOCT PLAUSIBLE: First part is correct.

Second part is plausible if the applicant does not understand the significance of the alarm being lit or not lit. In other words, if the applicant does not understand that there has to be sufficient inventory in the Containment Sump prior to swapover to prevent vortexing of the ND pumps, the second part is plausible. Additionally, the Containment Sump level alarms being LIT under these plant conditions is normal.

Answer D Discussion INCORRECT. See explanation above.

PLAUSIBLE: First part is plausible because the level is less than the level at which FWST makeup is required.

Second part is plausible if the applicant does not understand the significance of the alarm being lit or not lit. In other words, if the applicant does nt understand that there has to be sufficient inventory in the Containment Sump prior to swapover to prevent vortexing of the ND pumps, the second part is plausible. Additionally, the Containment Sump level alarms being LIT under these plant conditions is normal.

Basis for meeting the KA The applicant must understand the significance of the Containment Sump level alarms relative to plant conditions to know that the ND pumps must be stopped if FWST level decreases below a minumum level and sufficient inventory does not exist in the Containment Sump at the time of swapover to prevent vortexing of the ND pumps.

Basis for I-li Cog This is a higher cognitive level question because it requires multiple mental steps. First the applicant must analyze the data given to understand that NS pumps are not running (i.e. Containment pressure peaked at 2.5 PSIG). The applicant must then recall from memory that less than 250 inches with both CONT SUMP LEVEL GREATER THAN 2.5 FT alarms DARK requires tripping both ND pumps.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures):

1) The question can NOT be answered solely by knowing systems knowledge. Securing the ND Pumps if the potential for vortexing exists upon reaching the swapover point is not addressed in the limits and precautions or in the Systems Lesson Plan. Therefore, this is not systems level knowledge.

2) The question can NOT be answered by knowing immediate operator actions. There are no immediate actions associated with ES- 1.2.

3) The question can NOT be answered solely by knowing entry conditions for AOP or direct entry conditions for EOPs.

4) The question can NOT be answered solely by knowing the purpose, overal sequence of events, or overall mitigative strategy of the procedure.

5) This is detailed knowledge of a procedure diagnostic step that requires specific actions to be taken if conditions are not met. Therefore, this is SRO level knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided Learning Objectives:

1) EPE1004 L

Friday, May 14, 2010 Page 253 of 275

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2010 MNS SRO NRC Examination QUESTION 93 2593

References:

El Background Document KA_desc WEO3 WEO3 GENERICLIAbiIity to verif that the alarms are consistent with the plant conditions. (CFR: 41.10/ 43.5 /45.3 /45.12) 2.4.46 4O1-9 Comments: RemarkslStatus Friday, May 14, 2010 Page 254 of 275

Question 93

References:

From ES-1.2 Background Document:

STEP 6 Check I ETA and I ETB - ENERGIZED BY OFFSITE POWER PURPOSE: To ensure that the vital 41 60V AC busses are energized.

BASIS: If offsite power is available, station equipment should be aligned to the offsite source. If either vital bus is NOT energized from its offsite source, AP111A15500107 (Loss of Electrical Power) should be referenced to ensure the automatic loading of equipment on the bus (e.g., charging pumps, MD CA pumps, KC pumps, etc.). This AP also provides actions to realign offsite power to a vital bus when the offsite source becomes available, addresses maintaining DC busses, and other issues associated with a loss of power.

STEP 7 Place all Pzr heaters in manual and off.

PURPOSE: To turn off all Pzr heaters prior to restoring Pzr level in order to minimize NC heat input.

BASIS: This action, consistent with normal cooldown procedures, prevents Pzr heat inputs from being automatically initiated. This added heat would tend to keep the NC pressurized.

NOTE: If all NC pumps are off, the upper head region may void during NC System depressurization. This will cause Pzr level to rise rapidly.

PURPOSE: To alert the operator of possible void formation in the NC during the NC depressurization.

BASIS: As the NC system is depressurized, steam may form in the hotter regions on the NC system. Pzr level will rise rapidly as water displaced from these voided regions replaces steam in the pressurizer. If voiding occurs, the Pzr may fill with water within a few minutes. This note informs the operator of this condition so that the NC system depressurization can be stopped quickly to avoid a water solid pressurizer.

STEP 8 Check if ND pumps should be stopped: (CONTINUOUS ACTION)

PURPOSE: To stop the ND pumps if NC pressure is above their shutoff head to prevent damage to the pumps.

BASIS: Upon S/I initiation all safeguard pumps are started regardless of the possibility of high NC pressure with respect to the low-head S/I pump shutoff head. On low-head systems where the pump recirculates at low flow there is concern with long term operation at low flow rates. Shutdown of the pump when the NC pressure meets the criteria outlined in this step allows for future pump operability. However, if NC pressure goes below 286 psig the pumps will have to be manually restarted since no automatic signal is available.

Additional criteria for stopping ND pumps were added to the step. For some low temperature mode 3 scenarios (described in PIP M-04-5515), the existing ERG step would leave ND pumps running with suction on FWST. For these small break LOCA

events, NS does not actuate. If FWST level reaches 250 inches and inadequate sump level is indicated, ND pumps must be secured prior to auto swapover to prevent them from vortexing. 250 inches was selected to provide many minutes for operators to respond. This step will also energize ND discharge valves and allow using them to isolate ND if single failure occurs preventing securing of ND pump. As documented in PIP M-04-5115, corrective action 11, ND operation for 10 minutes is always enough to ensure core reflood for an event initiated in Mode 3. By the time 250 inches FWST level is reached, ND operation much longer than this is assured. The TSC is requested to help monitor FWST level, since there is no alarm at 250 inches.

STEP 9 Control intact SIC levels: (CONTINUOUS ACTION)

PURPOSE: To ensure adequate feed flow or SIG inventory for secondary heat sink requirements.

BASIS: The minimum feed flow requirement satisfies the feed flow requirement of the Heat Sink Status tree until level in at least one S/G is restored into the narrow range. Narrow range level is reestablished in all S/Gs to maintain symmetric cooling of the NC. The control range ensures adequate inventory with level readings on span.

STEP 10 Initiate NC System cooldown to Cold shutdown:

PURPOSE: To begin or continue a controlled NC cooldown to cold shutdown using a preferred or alternate method with a specified maximum cooldown rate.

BASIS: The objective of a controlled cooldown is to reduce the overall temperature of the NC coolant and metal to reduce the need for supporting plant systems and equipment required for heat removal. The maximum cooldown rate of 100°F/hr will preclude violation of the Integrity Status Tree thermal shock limits. The preferred steam release path is to the condenser to conserve inventory; however, atmospheric release is the stated alternative. The ND system may have been placed in RHR mode later in the procedure, and should be used to cool down the NC to cold shutdown.

STEP 11 Check NC subcooling based on core exit TICs GREATER -

THAN 0°F.

PURPOSE: To determine if the NC is subcooled so that subsequent actions dependent upon subcooling can be performed.

BASIS: If NC subcooling can be verified, the LOCA is most likely small and controllable, i.e., S/I flow equals or exceeds break flow. Subsequent steps that may be allowed include deliberate NC depressurization, NC pump restart, and S/I flow reduction. If subcooling is inadequate the operator is directed to increase S/I flow to restore subcooling.

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2010 MNS SRO NRC Examination QUESTION 94 2594 J_desc GEN2. 1 Conduct of Operations D Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc. (CFR: 41.10/ 43.2) 2.1.4 Unit I is operating at 100% RTP.

An active licensed STA may assume the duties of the Control Room Supervisor provided the CRS or relief SRD is available to return to the control room within (1) AND the periods during which the STA assumes SRO duties do not exceed (2) in duration.

Which ONE (1) of the following completes the statement above?

A. 1. 10 minutes

2. 15 minutes B. 1. 15 minutes

2. 10 minutes C. 1. 15 minutes

2. 15 minutes

, D. 1. 10 minutes

2. 10 minutes Friday, May 14, 2010 Page 255 of 275

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 94 t3eneral Discussion Technical Specifications allows the Shift Technical Advisor to assume the control room command function and perform the duties of the control room SRO in Modes 1, 2, 3, and 4 during periods when the CRSRO and the relief SRO are required to be absent from the control room.

However, the following requirements must be met:

- The STA must hold an SRO license for the unit.

- The CRSRO or relief SRO must be available to return to the control room within 10 minutes.

- The periods during which the STA may perform the control room SRO duties may not exceed 15 minutes in duration or a total of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for the entire shift.

Answer A Discussion CORRECT.

Answer B Discussion Incorrect. Plausible if the candidate confuses the time for the CRSRO or relief SRO to return to the control room with the allowable duration of the relief by the STA.

Answer C Discussion Incorrect. Plausible if the candidate confuses the time for the CRSRO or relief SRO to return to the control room with the allowable duration of relief by the STA.

Answer D Discussion Incorrect. Plausible if the candidate confuses the time for the CRSRO or relief SRO to return to the control room with the allowable duration of the relief by the STA.

Basis for meeting the KA KA is matched because the candidate must understand the control room manning requirements for the individual fulfilling the control room command function.

Basis for Hi Cog lasis for SRO only fhis question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for screening questions linked to 10CFR55.43(b)(1 & 2) (Tech Specs):

1) This question can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs. These requirements are in 5.1.2 which has no action statements.

2) This question can NOT be answered by knowing information listed above-the-line. These are administrative requirements. There is no above-the-line knowledge.

3) This question can NOT be answered by knowing the TS Safety Limits or their bases. This is TS 5.1.2. not TS Safety Limits.

4) This question requires the applicant to have knowledge of TS administrative requirements contain in Section 5 of Tech Specs. This is SRO level knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Memory BANK 2009 MNS SRO Exam Development References Student References Provided Learning Objective:

1) OP-MC-ADM-OMP. Obj 3

References:

1) Technical Specification 5.1.2, amendment 213 and 194 KA KA_desc GEN2. 1 Conduct of Operations D Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solooperation, maintenance of active license status, 10CFR55, etc. (CFR: 41.10 / 43.2) 2.1.4 401-9 Comments: RemarkslStatus Friday, May 14, 2010 Page 256 of 275

FOR REVIEW ONLY DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 94 Friday, May 14, 2010 Page 257 of 275

Question 94

References:

From T.S. 5.1.2:

Reportability 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1.1 The Station Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

5.1.2 The Control Room Senior Reactor Operator (CRSRO) shall be responsible for the control room command function. During any absence of the CRSRO from the control room while the unit is in MODE 1 2, 3, or 4, an individual [other than the Shift Technical Advisor (STA)] with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. During any absence of the CRSRO from the control room while the unit is in MODE 5 or 6, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.

On occasion when there is a need for both the CRSRO and the relief SROto be absent from the control room in MODE 1,2,3, or4 the STA shall be allowed to assume the control room command function and serve as the SRO in the control room provided that:

a. the CRSRO or the relief SRC is available to return to the control room within 10 minutes.

b. the assumption of SRO duties by the STA is limited to periods not in excess of 15 minutes duration and a total time not to exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during any shift, and

c. the STA has a SRO license on the unit.

MoGuire Units I and 2 5.1-1 Amendment Nos. 213 / 194

Question 94 Parent Question:

Examination Outline Cross-reference: Level RO SRO x

Tier# 3 Final Group K/A# G2.1.5 Importance Rating 3.9 Conduct of operations Ability to locate and use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

Proposed Question: SRO 86 1 Pt Unit 1 is operating at 100% RTP.

Under which ONE (1) of the following conditions may an active licensed STA assume the duties of the Control Room Supervisor?

The CRS or relief SRO is available to return to the control room within (1)

AND the periods during which the STA assumes SRO duties do not exceed (2) in duration.

A. (1) 15 minutes (2) 15 minutes B. (1) 15 minutes (2) 10 minutes C. (1) 10 minutes (2) 15 minutes D. (1) 10 minutes (2) 10 minutes Proposed Answer: C

Explanation (Optional):

Technical Specifications allows the Shift Technical Advisor to assume the control room command function and perform the duties of the control room SRO in Modes 1, 2, 3, and 4 during periods when the CRSRO and the relief SRO are required to be absent from the control room. However, the following requirements must be met:

The STA must hold an SRO license for the unit.

The CRSRO or relief SRO must be available to return to the control room within 10 minutes.

The periods during which the STA may perform the control room SRO duties may not exceed 15 minutes in duration or a total of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for the entire shift.

A. Incorrect: See explanation above. Plausible if the candidate confuses the time for the CRSRO or relief SRO to return to the control room with the allowable duration of the relief by the STA.

B. Incorrect: See explanation above. Plausible if the candidate confuses the time for the CRSRO or relief SRO to return to the control room with the allowable duration of the relief by the STA.

C. Correct.

D. Incorrect: See explanation above. Plausible if the candidate confuses the time for the CRSRO or relief SRO to return to the control room with the allowable duration of the relief by the STA.

Technical Reference(s) Technical Specification 5.1.2, (Attach if not previously provided) amendment 213 and 194 (Including version or revision #)

Proposed references to be provided to applicants during examination: None Learning Objective: OP-MC-ADM-OMP, Obj 3 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 43.5 Comments:

Conduct of operations Ability to locate and use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

KA is matched because the candidate must understand the control room manning requirements for the individual fulfilling the control room command function.

This is an SRO Only question linked to 10CFR55.43(b)(2), Tech Specs. This questions can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Spec or TRM action statements. It can NOT be answered by knowing the LCO/TRM information listed above-the-line (since this is an Administrative Control). It can NOT be answered by knowing Tech Spec Safety Limits or their basis. The candidate must apply requirements from Section 5.0, Administrative Controls of Technical specifications. Requirements in Section 5.0 are NOT expected knowledge for ROs.

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2010 MNS SRO NRC Examination QUESTION 95 2595 KA_desc Conduct of Operations: Ability to coordinate personnel activities outside the control room. (CFR: 41.10/45.5/45.12/

2.1.8 Given the following conditions on Unit 1:

The unit is in a refueling outage

Fuel movement is in progress

A leak has developed which has caused level to drop in the spent fuel pool

The Spent Fuel Pool Level Low computer alarm has actuated In accordance with AP-40 (Loss of Refueling Canal Level), which ONE (1) of the following describes the FIRST action directed by the CRS to mitigate the current conditions?

A. Place the weir gate in position and inflate the seals.

B. Begin makeup to the pool from the Boric Acid Tank.

C. Move the fuel transfer cart to the reactor side and close 1KF-122 (Fuel transfer tube block valve).

D. Move the fuel transfer cart to the spent fuel (pit) side and close 1 KF-122 (fuel transfer tube block valve).

Friday, May 14, 2010 Page 258 of 275

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2010 MNS SRO NRC Examination QUESTION 95 2595 Jeneral Discussion In accordance with AP-40 the first action which will be directed by the CRS is to move the fuel transfer cart to the spent fuel pit side and close IKF-122.

Answer A Discussion INCORRECT. See explanation above.

PLAUSIBLE: This answer is plausible because this would be the correct answer if IKF-122 could not be closed. However, the first attempt is to close 1KF-122.

Answer B Discussion fNCORRECT. See explanation above.

PLAUSIBLE: This answer is plausible because the Operators are directed to make upt o the spent fuel pool in AP-40. However, the first action is to attempt to isolate the spent fuel pooi from the refueling canal the preserve the water that is in the spent fuel pool. Also, makeup to the spent fuel pooi is not normally done from the BAT.

Answer C Discussion INCORRECT. See explanation above.

PLAUSIBLE: This answer is plausible if the applicant does not recall which side of the transfer tube the fuel transfer cart has to be located to close the block valve (1KF-122).

Answer DDiscussion CORRECT. See explanation above.

Basis for meeting the KA The KA is matched because the applicant must have knowledge of local operator actions outside of the control room to be able to coordinate those activities.

Basis for Hi Cog iasis for SRO only

This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures):

I) The question can NOT be answered by knowing systems knowledge alone. This is detailed procedure content knowledge from AP-40 and AP 41.

2) The question can NOT be answered by knowing immediate Operator actions. There are no immediate operator actions associated with AP-40 orAP-4l.

3) The question can NOT be answered by knowing AOP or EOP entry conditions. Knowledge of AP-40 entry conditions will not enable the applicant to correctly answer this question.

4) The question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of AP-40 or AP-4 1.

5) The question requires the applicant to assess plant conditions and then prescribing a procedure or section of a procedure to mitigate the consequences of the event. Specific to this event, initial entry would be into AP-4l (Loss of SFP Cooling or Level). However, since 1KF-122 is open the operator is directed out of AP-4 I and into AP-40 (Loss of Refueling Canal Level) where they are directed to perform the appropriate actions.

Job Level Cognitive Level QuestionType Question Source SRO Memory BANK MNS Exam Bank Question FHFCNO14 Development References References Provided Learning Objective:

1)

References:

Lesson Plan OP-MC-FH-FC Section 3.2.2

) AP-40 Friday, May 14, 2010 Page 259 of 275

FOR REVIEW ONLY DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 95

( 2.1.8 coordinate personnel activities outside the control room. (CFR: 41.10/45.5/45.12/

L 401 -9 Comments: RemarkslStatus Friday, May 14, 2010 Page 260 of 275

Question 95

References:

From Lesson Plan OP-MC-FH-FC Section 3.2.2:

The Symptoms include:

EMF36 UNIT VENT GAS HI RAD alarm

EMF38 CONTAINMENT PART HI RAD alarm

EMF39 CONTAINMENT GAS HI RAD alarm

EMF4O CONTAINMENT IODINE HI RAD alarm

EMF42 FUEL BLDG VENT HI RAD alarm

EMF16 CONTAINMENT REFUELING BRIDGE alarm (2 EMF3 on Unit 2) -

EMF17 SPENT FUEL BLDG REFUEL BRDG alarm (2 EMF4 on Unit 2)-

Gas bubbles originating from the damaged assemblies

Visible evidence of damage with the potential of radioactive releases Operator Actions CAUTION Damage to the rubber Reactor Vessel Cavity Seal may occur if an assembly is dropped on or near it.

Announce on page. If in containment, evacuate containment, assemble in contaminated change room and refer to RP/01A1570011 1, Conducting a Site Assembly, Site Evacuation, or Containment Evacuation. Isolate containment: stop VP fans, ensure VP valves close, stop any VQ release, ensure equipment hatch closed, ensure one airlock door closed, dispatch Operator to move conveyor to Spent Fuel Pool Building, dispatch Operator to close KF-122. If high containment radiation exists, place Aux Carbon Filters in service per OP. Place Refueling Cavity in purification per OP.

If in Spent Fuel Building, evacuate Spent Fuel Pool area, assemble in contaminated change room. Isolate Spent Fuel Pool area: Check if VF EXH BYP DAMPER closed lite lit, and if not, place its control switch to CLOSE, and close the doors to the Spent Fuel Pool area. Ensure KF purification loop in service per OP.

Refer to RPIOIAI5700IOO, Classification of Emergency.

3.2.2 AP111A15500140, LOSS OF REFUELING CANAL LEVEL The purpose is to provide actions in the event of loss of water in the refueling canal.

The Symptoms include:

Spent Fuel Pool Level Low computer alarm

Decreasing level in refueling canal

Incore Inst Room Sump Hi Level alarm

EMF16 CONTAINMENT REFUELING BRIDGE alarm (2 EMF3 on Unit 2) -

EMF17 SPENT FUEL BLDG REFUEL BRDG alarm (2 EMF4 on Unit 2)-

Operator Actions NOTE Any available core location may be used when lowering a fuel assembly during emergency conditions.

If fuel movement is in progress: lower any assembly in the reactor building crane to fully down in the core, any assembly in the spent fuel crane to fully down, and any assembly in the upender to fully down. If they wont lower otherwise, manually release the brake and hand crank the hoist down. NOTE: The sequence for lowering the hoist manually should be to put the emergency handwheel on the end of the hoist motor, hold it steady, while another person screws in the brake release (star shaped knob on a threaded stud) which when threaded in forces the brake disengaged. Care should be taken to remove the handwheel before electric operation of the hoist motor. The upender is similar. The bridge and trolly brake release is a lever, otherwise similar. Dispatch Operator to locally move fuel transfer cart to the spent fuel (pit) side. Stop FWST Pump and close FW-13, and dispatch Operator to locally close KF-122. If KF-122 cannot be closed, then notify RP to begin surveys, consider installing the weir gate, and isolate the Spent Fuel Building (VF in filter mode and doors closed). Evacuate nonessential personnel from containment and Spent Fuel Building.

Try to identify and correct the cause of decreasing level. Verify seal integrity and air pressure to the Rx Vessel cavity seal and the Rx Vessel nozzle inspection port seals, and if not, reestablish VI to seals. Dispatch an Operator to locally ensure the Refueling Cavity Drains are closed. Check the SIC Nozzle Dams. Refer to API1 9, Loss of ND or ND System Leakage, while continuing with this procedure.

Makeup to the canal per OP/1/A/6200/13. CAUTION: Makeup to the SEP could dilute NC system boron concentration.

Monitor the Spent Fuel Pool level. If it gets to minus two feet, stop the KF Pump and turn off the lights. Initiated makeup per OP. If pool level low enough for radiation hazard, makeup from RN.

Ensure Containment Integrity with equipment hatch and airlock doors closed. If time permits, turn off canal underwater lights before they become uncovered. If necessary due to increasing radiation levels, consider using ND or NS to transfer water from the containment sump to the FWST for additional makeup capability.

Refer to RPIOIAI5700IOO, Classification of Emergency.

From AP-40:

MNS LOSS OF REFUELING CAVITY LEVEL PAGE NO.

AP.iA/5500/40 2 of i Rev. 7 UNIT 1 CNLEKEDTED RESONE PESONSE JOP OBTMNED B $ymp.toms

SPENT FUEL POOL LEVEL LOW computer alarm

Level in refueling cavity going down

INCORE INST ROOM SUMP HI LEVEL alarm

IEMF-16 CONTAINMENT REFUELING BRDG alarni

1EMF-17 SPENT FUEL BUILDING BRDG alarm.

C. OperatQr Actions

1. Announce occurrence on page.

2. -

-I-Check FUEL MOVEMENT IN Perform the following:

PROGRESS.

a. jf any radio active component is being handled ri the spent fuel pool or refueling cavity. THEN have fuel handling crew lower component to fully down.

b. IF cavity level is dropping more than one inch per minute, AND 1 FW-27A (Unit 1 FWST to ND Pumps Isol) is open, THEN initiate makeup PER Enclosure 3 (Refueling Cavity Makeup Using ND Pump) while continuing in this AR

c. GOTOStep4.

MNS LOSS OF REFUELING CAVITY LEVEL PAGE NO.

AP/1A,.5500140 3of 8 UNIT 1 Rev. 7 DT:cN.ExE.:TEz RESCNE RESOSE SOT OBTISED NPTE Any available core location may be used when lowering a fuel assembly duriii emergency conditions.

3. Contact fuel handting SRO to have fuel handling crew perform the following:

a. Lower any fuel assembly in the reactor a. Release broke and hand crank hoist bLiilding manipulator crane to fully down down.

in the core.

4

b. Lower any fuel assembly in the spent

b. Release brake and hand crank hoist fuel manipulator crane to fully down. down.

4

c. Lower any fuel assembly in either a. Release brake and hand crank upender upender to fully down. down.

d. Move fuel transfer cart to the spent fuel d. Release brake and hand crank transfer (Pit) side. cart to spent fuel (Pit) side.

e. Lower any radioactive component in the e. Perform the following:

spent fuel pool or refueling cavity to fully down.

Reinstall component.

I OR

. Place component as far below the water as safely possible.

4. WHEN fuel transfer cart is in the spent fuel bldg, THEN dispatch 2 operators to CLOSE I KF-I 22 (Unit I: Fuel Transfer Tube Isol) (spent fuel bldg. 780, PP-SI, top of fuel pool at south east corner).

5. Notify Containment C(osure Coordinator to initiate containment closure PER PT111N42001002 C (Containment Closure).

Question 95 Parent Question:

FHFCNO14 1 Pt(s) Given the following conditions:

Unit 1 is in a refueling outage.

Fuel movement is in progress.

A leak has developed which has caused level to drop in the spent fuel pool.

The Spent Fuel Pool Level Low computer alarm has actuated.

Pool was initially at normal level and area radiation at 7 mrem/hr.

After 20 minutes the pool level has decreased further and area radiation is 18 mrem/hr.

Which one (1) of the following describes the operator response to the current conditions?

A. Begin makeup to the pool from the Boric Acid Tank. to restore level.

B. Move the fuel transfer cart to the reactor side and close 1 KF-122 (Fuel transfer tube block valve).

C. Move the fuel transfer cart to the spent fuel (pit) side and close I KF 122 (fuel transfer tube block valve).

D. Place the weir gate in position and inflate the seals.

Answer 6 Answer: C MISCINFO: SRO Only SOURCE: SR092

REFERENCES:

AP/1/A/5500/40, p. 2,3 LESSON: OP-MC-FH-FCB TASK:

p. 59, 60 OBJECTIVE: 15C TIME:

K/A: 000036G010 (3.7/3.8) DATE: SROEXAM1992

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2010 MNS SRO NRC Examination QUESTION_96 KA TKA_desc GEN2.2 Equipment ControlElAbility to apply Technical Specifications for a system. (CFR: 41.10/ 43.2 / 43.5 /45.3) 2.2.40 While the Unit is in MODE 1 Shutdown Margin is verified by (1)

The basis for ensuring Shutdown Margin is maintained in MODE 1 is to protect againsttheeffectsof (2)

Which ONE (1) of the following completes the statement above?

A. 1. performing a SDM calculation

2. an ejected rod B. 1. checking control rods above insertion limits

2. an ejected rod C. 1. performing a SDM calculation

2. a steam line break D. 1. checking control rods above insertion limits

2. a steam line break Friday, May 14, 2010 Page 261 of 275

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2010 MNS SRO NRC Examination QUESTION 96 259 eneraI Discussion Per TS 3.1.1 bases: During power operation, SDM control is ensured by operating with the shutdown banks fully withdrawn and the control banks within the limits of LCO 3.1.6, Control Bank Insertion Limits.

Per TS 3.1.6 bases: The shutdown and control bank insertion and alignment limits, AFD, and QPTR are process variables that together characterize and control the three dimensional power distribution of the reactor core. Additionally, the control bank insertion limits control the reactivity that could be added in the event of a rod ejection accident, and the shutdown and control bank insertion limits ensure the required SDM is maintained.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE: Part I Student may recall that a SDM verification is required in Mode 1 but misinterpret what method is required per TS Basis.

The actual calculation would be performed in Mode 1 if there was an issue with Rod Insertion Limits or an inoperable control rod.

Part 2 is correct and therefore plausible.

Answer B Discussion ORRECT: See explanation above.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE: Part 1 Student may recall that a SDM verification is required in Mode 1 but misinterpret what method is required per TS Basis.

The actual calculation would be performed in Mode 1 if there was an issue with Rod Insertion Limits or an inoperable control rod.

Part 2 is plausible because a SLB is discussed extensively in the SDM basis (TS 3.1.1) but may misapply this consideration to apply to Mode 1 as well as Mode 2.

Answer D Discussion CORRECT: See explanation above.

PLAUSIBLE: Part 1 is correct and therefore plausible.

Part 2 is plausible because a SLB is discussed extensively in the SDM basis (TS 3.1.1) but may misapply this consideration to apply to Mode 1 as well as Mode 2.

Basis for meeting the KA For this question, the applicant must have knowledge of surveillances related to meeting SDM. There is no TS for SDM in MODE 1. However, SDM met be met at all times, even in MODE 1. The applicant demonstrates the apply Technical Specifications by demonstrating an in-depth understanding of how we verify SDM is being maintained in MODE 1 (insertion limits) and in MODES 2 and below (by SDM calculation).

Therefore, the KA is matched.

Basis for Hi Cog Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for screening questions linked to IOCFR55.43(b)(2) (Tech Specs):

1) This question can NOT be answered by knowing less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs. This question is related to knowledge of TS surveillance requirements (for SDM surviellance in both MODE 1 and MODE 2 and below) that are greater than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months specifications.

2) This question can NOT be answered by knowing information listed above-the-line. Surveillance requirements are below-the-line knowledge.

3) This question can NOT be answered by knowing the TS Safety Limits or their bases. TS 3.1.1 SDM and TS 3.1.6 Control Bank Insertion Limits.

4) This question requires the applicant to have knowledge of the application of TS Surviellance requirements and the basis for those surveillance requirements. Therefore, this is SRO level knowledge.

This Question is linked to 10CFR55.43(b)(2) for SRO only because it requires the applicant to recall information contained in the TS basis for SDM (TS 3.1.1), specifically what limiting accident the LCO is required to address and the mode of applicability for the specific accident. The applicant is further required to recall information contained in the basis for TS 3.1.6 (Control Bank Insertion Limits) associated with how the Friday, May 14, 2010 Page 262 of 275

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2010 MNS SRO NRC Examination QUESTION 96 2596 iification of rod insertion limits is used to ensure the requirements of TS 3.1.1 are met in Mode 1.

Job Level Cognitive Level QuestionType Question Source SRO Memory BANK 2009 AUDIT Q96 (Bank 1396)

Development References Student References Provided

References:

1)TS3.1.1 and Basis

[S3.1.6andBasis KA KAdesc GEN2.2 Equipment ControlElAbility to apply Technical Specifications for a system. (CFR: 41.10/43.2 / 43.5 / 45.3) 2.2.40 401-9 Comments: RemarkslStatus Friday, May 14, 2010 Page 263 of 275

SDM B 3.1.1 Question 96

References:

From Tech Spec 3.1.1 (SDM) Basis:

B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)

BASES BACKGROUND According to GDC 26 (Ref. 1), the reactivity control systems must be redundant and capable of holding the reactor core subcritical when shut down under cold conditions. Maintenance of the SDM ensures that postulated reactivity events will not damage the fuel.

SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs). As such, the SDM defines the degree of subcriticality that would be obtained immediately following the insertion or trip of all shutdown and control rods, assuming that the single rod cluster control assembly of highest reactivity worth is fully withdrawn.

The system design requires that two independent reactivity control systems be provided, and that one of these systems be capable of maintaining the core subcritical under cold conditions. These requirements are provided by the use of movable control assemblies and soluble boric acid in the Reactor Coolant System (RCS). The Rod Control System can compensate for the reactivity effects of the fuel and water temperature changes accompanying power level changes over the range from full load to no load. In addition, the Rod Control System, together with the boration system, provides the SDM during power operation and is capable of making the core subcritical rapidly enough to prevent exceeding acceptable fuel damage limits, assuming that the rod of highest reactivity worth remains fully withdrawn. The soluble boron system can compensate for fuel depletion during operation and all xenon burnout reactivity changes and maintain the reactor subcritical under cold conditions.

During power operation, SDM control is ensured by operating with the shutdown banks fully withdrawn and the control banks within the limits of LCO 3.1.6, Control Bank Insertion Limits. When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration.

McGuire Units 1 and 2 B 3.1.1-1 Revision No. 73

SDM B 3.1.1 BAS ES APPLICABLE The minimum required SDM is assumed as an initial condition in safety SAFETY ANALYSES analyses. The safety analysis (Ref. 2) establishes an SDM that ensures specified acceptable fuel design limits are not exceeded for normal operation and AOOs, with the assumption of the highest worth rod stuck out on a reactor trip.

The acceptance criteria for the SDM requirements are that specified acceptable fuel design limits are maintained. This is done by ensuring that:

a. The reactor can eventually be made subcritical from all operating conditions, transients, and Design Basis Events;

b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits (departure from nucleate boiling ratio (DNBR), fuel centerline temperature hmits for AOOs and 280 cal/gm energy deposition for the djcon ent]and

c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

One limiting accident for the SDM requirements is based on a main steam line break MSL) Moas described in the accident analysis (Ref. 2). The increased steam flow resulting from a pipe break in the main steam system causes an increased energy removal from the affected steam generator (SG), and consequently the RCS. This results in a reduction of the reactor coolant temperature. The resultant coolant shrinkage causes a reduction in pressure. In the presence of a negative moderator temperature coefficient, this cooldown causes an increase in core reactivity. As RCS temperature decreases, the severity of an MSLB decreases. The most limiting MSLB, with respect to potential fuel damage before a reactor trip occurs, is a break of a main steam line upstream of the Main Steam Isolation Valve initiated at the end of core life. The positive reactivity addition from the moderator temperature decrease will terminate when the affected SG boils dry, thus terminating RCS heat removal and cooldown. Following the MSLB, a post-trip return-to-power may occur; however, no fuel damage occurs as a result of the post-trip return-to-power, and THERMAL POWER does not violate the Safety Limit (SL) requirement of SL 2.1.1.

A potentially more limiting MSLB accident could occur for a steam line break outside containment when in Mode 3 with the low pressurizer pressure signal for safety injection actuation blocked. In this scenario, feedwater would not automatically isolate and the peak heat fluxes associated with the return-to-power may increase to values significantly McGuire Units 1 and 2 B 3.1.1-2 Revision No. 73

SDM B 3.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) greater than those in the accident analysis (Ref. 2). Therefore, when safety injection is blocked, administrative controls on boron concentration are required to prevent a return-to-power following a steam line break.

In addition to the limiting MSLB transient, the SDM requirement must also protect against:

a. Inadvertent boron dilution;

b. An uncontrolled rod withdrawal from subcritical or low power condition; and

c. Roejcti.

Each of these events is discussed below.

In the boron dilution analysis, the required SDM defines the reactivity difference between an initial subcritical boron concentration and the corresponding critical boron concentration. These values, in conjunction with the configuration of the RCS and the assumed dilution flow rate, directly affect the results of the analysis. This event is most limiting at the beginning of core life, when critical boron concentrations are highest.

Depending on the system initial conditions and reactivity insertion rate, the uncontrolled rod withdrawal transient is terminated by either a high power level trip or a high pressurizer pressure trip. In all cases, power level, RCS pressure, linear heat rate, and the DNBR do not exceed allowable limits.

The ejection of a control rod rapidly adds reactivity to the reactor core, causing both the core power level and heat flux to increase with corresponding increases in reactor coolant temperatures and pressure.

The ejection of a rod also produces a time dependent redistribution of core power. SDM satisfies Criterion 2 of 10 CFR 50.36 (Ref. 3). Even though it is not directly observed from the control room, SDM is considered an initial condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of accident analysis assumptions.

Transients which are made less severe by the rapid insertion of control rod negative reactivity are also affected by the magnitude of the SDM limit. This is because the safety analyses assume a change in the rate of insertion of this negative reactivity when the SDM limit is reached. While the SDM is less than the limit value, the negative reactivity from the McGuire Units 1 and 2 B 3.1.1-3 Revision No. 73

SDM B 3.1.1 BASES APPLICABLE SAFETY ANALYSES (continued) control rods is assumed to be inserted as quickly as the rod worth vs.

time curves shown in Reference 5. When the SDM limit value is reached, the rate of negative reactivity insertion is decreased so that it is only fast enough to compensate for any positive reactivity insertion, e.g., from the cooling of the fuel and moderator (which normally have negative temperature coefficients). This methodology is conservative in that it does not take credit in the safety analyses, even temporarily, for a SDM greater than the limit value.

LCD The MSLB (Ref. 2) and the boron dilution (Ref. 4) accidents are the most limiting analyses that establish the SDM value of the LCD. For MSLB accidents, if the LCD is violated, there is a potential to exceed the DNBR limit and to exceed 10 CFR 100, Reactor Site Criteria, limits (Ref. 5).

For the boron dilution accident, if the LCD is violated, the minimum required time assumed for operator action to terminate dilution may no longer be applicable.

APPLICABILITY In MDDE 2 with keff < 1.0 and in MDDES 3, 4, and 5, the SDM requirements are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above. In MDDE 6, the shutdown reactivity requirements are given in LCD 3.9.1, Boron Concentration. In MDDES 1 and 2 with keff 1.0, SDM is ensured by complying with LCD 3.1.5, Shutdown Bank Insertion Limits, and LCD 3.1.6.

McGuire Units 1 and 2 B 3.1.1-4 Revision No. 73

SDM B 3.1.1 BASES From Tech Spec 3.1.6 (Rod Insertion Limits) Basis:

BACKGROUND (continued)

The control banks are used for precise reactivity control of the reactor. The positions of the control banks are normally controlled automatically by the Rod Control System, but can also be manually controlled. They are capable of adding reactivity very quickly (compared to borating or diluting).

The power density at any point in the core must be limited, so that the fuel design criteria are maintained. Together, LCO 3.1.4, LCO 3.1.5, Shutdown Bank Insertion Limits, LCO 3.1.6, LCO 3.2.3, AXIAL FLUX DIFFERENCE (AFD), and LCO 3.2.4, QUADRANT POWER TILT RATIO (QPTR), provide limits on control component operation and on monitored process variables, which ensure that the core operates within the fuel design criteria.

The shutdown and control bank insertion and alignment limits, AFD, and QPTR are process variables that together characterize and control the three dimensional power distribution of the reactor core. Additionally, the control bank insertion limits control the reactivity that could be added in the event of a rod ejection accident, and the shutdown and control bank insertion limits ensure the required SDM is maintained.

Operation within the subject LCO limits will ensure that fuel cladding failures that would breach the primary fission product barrier and release fission products to the reactor coolant in the event of a loss of coolant accident (LOCA), ejected rod, or other accident requiring termination by a Reactor Trip System (RTS) trip function are no more than those predicted in and allowed by the safety analyses.

APPLICABLE The shutdown and control bank insertion limits, AFD, and QPTR LCOs SAFETY ANALYSES are required to prevent power distributions that could result in excessive fuel cladding failures in the event of a LOCActed or other accident requiring termination by an RTS trip function.

The acceptance criteria for addressing shutdown and control bank insertion limits and inoperability or misalignment are that:

a. There be no violations of:

McGuire Units I and 2 B 3.1.1-5 Revision No. 73

SDM B 3.1.1 BASES

1. specified acceptable fuel design limits, or

2. Reactor Coolant System pressure boundary integrity; and

b. The core remains subcritical after accident transients (e flne break acciden

/

This statement provides the plausibility for the distracters containing a steam line break as the basis for ensuring SDM.

McGuire Units I and 2 B 3.1.1-6 Revision No. 73

SDM B 3.1.1 BASES McGuire Units 1 and 2 B 3.1.1-7 Revision No. 73

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2010 MNS SRO NRC Examination QUESTION 97 2 KA_desc GEN2.2 Equipment ControlL/Knowledge of the process for making changes to procedures. (CFR: 41.10/43.3 /45.13) 2.2.6 --____________________

Given the following conditions on Unit 1:

The unit is in MODE 5 preparing for a unit startup after refueling

You are the Unit I Control Room Supervisor

A Temporary Test procedure is being run on the lB Boric Acid pump

The OATC points out that several steps in the TT procedure should be concurrent verification steps to be consistent with similar steps in other test procedures In accordance with NSD 703 (Administrative Instructions for Technical Procedures) the change to the Temporary Test Procedure should be processed as a (1) change and a 10 CFR50.59 Evaluation in accordance with NSD 228 (Applicability Determination) (2)

Which ONE (1) of the following completes the statement above?

A. 1. minor

2. is required B. 1. major

2. is required C. 1. minor

2. is NOT required D. 1. major

2. isNOTrequired Friday, May 14, 2010 Page 264 of 275

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2010 MNS SRO NRC Examination QUESTION 97 eneral Discussion In accordance with NSD 703 (Administrative Instructions for Technical Procedures) Section 703.4 (Criteria For Procedure Revisions and Changes) Step 4.4.3.e. on of the examples of changes that fit the definition of a minor procedure change is Add/delete inspection/verification signatures (e.g. QC Hold Point, Concurrent Verification). Therefore, this change is a minor procedure change.

In accordance with NSD 703 Section 7.2 (Review of Minor Procedure Changes) Step 5 An Applicability Determination (NSD 228) is NOT required for a minor procedure change. A 10 CFR5O.59 review is part of the Applicability Determination.

Answer A Discussion INCORRECT. See explanation above.

PLAUSIBLE: Part 1 is correct.

Part 2 is plausible if the applicant does not recall that an Applicability Determination (and 50.59 Review) is not required for minor procedure changes. It is plausible for an applicant to conclude that an Applicability Determination in accordance with NSD 228 is required regardless of whether the change is a minor or major revision. Part of an Applicability Determination is deciding whether a 50.59 review is required. If the

[pplicant concludes that an Applicability Determination is required they might also mistakenly conclude that a 50.59 Evaluation is also required.

Answer B Discussion NCORRECT. See explanation above.

PLAUSIBLE: Part 1 is plausible if the applicant does not recall the requirements of NSD 703 regarding the difference between a major and minor procedure change.

Part 2 is plausible because an Applicability Determination in accordance with NSD 228 is always required for a major revision and part of the Applicability Determination is deciding whether a 50.59 Evaluation is required. It is plausible that the applicant may mistakenly conclude that this means a 50.59 Evaluation needs to be performed.

Answer C Discussion ORRECT. See explanation above.

Answer D Discussion INCORRECT. See explanation above.

PLAUSIBLE: Part 1 is plausible if the applicant does not recall the requirements of NSD 703 regarding the difference between a major and minor procedure change.

Part 2 is correct. While an applicability determination is required for a major change, a 50.59 review may or may not be required for the change.

If a procedure is on the exclusion list for 50.59 reviews, it is not required for a major revision.

Basis for meeting the KA IKA is matched because the applicant must have knowledge of the Fleet Procedure requirements regarding changes to technical procedures.

Basis for Hi Cog Basis for SRO only This question meets the following examples for an SRO only question as described in the Clarification Guidance for SRO-only Questions Rev 1 dated 03/11/2010 for questions linked to 1 OCFR55.43(b)(3) (Facility licensee procedures required to obtain authority for design and operating changes in the facility):

10 CFR 50.59 screening and evaluation process

Processes for changing the plant or plant procedures Job Level Cognitive Level QuestionType Question Source SRO Memory NEW

)evelopment References Student References Provided Learning Objectives:

1) ADM-OP #2

References:

1) NSD 703 Administrative Instructions for Technical Procedures Friday, May 14, 2010 Page 265 of 275

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2010 MNS SRO NRC Examination QUESTION 97 2597 KA_desc - -____________________________________

GEN2.2 Equipment Contro1üIow1edge of the process for making changes to procedures. (CFR: 41.10/43.3 / 45.13) 2.2.6 401-9 Comments: RemarkslStatus Friday, May 14, 2010 Page 266 of 275

Question 97

References:

From NSD 703:

VERIFY HARD COPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USE NSD 703 Nuclear Policy Manual Volume 2

2. If the procedure change or revision does not alter the results. requirements. or methods by which a procedure is performed, go to 703.5 for processing a minor procedure revision or 7017 for processmg a minor procedure change.

3. The list below provides some examples of changes that fit the definition of a minor procedure rensiois or minor procedure change:

Note: TIns list is not intended to be all inclusive.

a. Incorporatme previously approved changes.

b. Correct editorial errors (e.g.. misspelled words. granimatical errors.

typographical esrors).

c. During the certification process. if the electroinc files have formatting differences from the approved version (e.g.. line endings. gage endings. word wraps).

it Correct, delete, or add information (e.g. numbering or references to steps.

pages. enclosures, or procedures. work location in&srmation. references to other documents. notes. cautions, warnings):

Adding. deleting, or correcting references to docunsents that are no longer applicable (e.g.. Site Directrve deleted in favor of an NSD}.

Changing the assigned part seiluenee munthers but the referenced part does not change (e.g.. Stock Code Numbers).

Changing the pan number where an acceptable ssibstitute has been identified under the Acceptable Substitute Program.

e. Correct. delete. or add nontechnical or administrative actions.

L Addidtiete mspecuon venficanon ss nalures (e.g.. QC hold Point. Concurrent Vrnfication)

g. Modify the format of a step or section. but not change the results:

Rephrasing a step. without changing its scope or results. to clarify.

Repinasiug to avoid ambiguous wording (e.g.. moving from general to specific).

Changing smits in a procedure data sheet (e.g.. Scale face as marked in inches. but the procedure specified readings in percent. Rather than requiring the teclmiciais to perform the conversion each time the procedure is performed, change the procethre to reflect field coiiflgsirarion.)

Ii. Reflect changes in admnsisirative work practices that are not commitment items:

Dividing Test Equipment into Test Eqsnpment and Other Equipment.

Deletusg Verify to Control Copy step (now on P PR).

Deleting step for Supervisor to NA steps before beguining procedure (now covered in NSD 704).

Deletiisg or adding step to Get key to open cabinet (when cabinet is no longer locked or is being locked).

Document changes initiatedisubstantiated by other processes evaluated in accordance with NSD 228 (Applicability Determination):

Equipment nunaber changes evaluated snider the engineen.isg change process.

12 REVISION 29 VERIFY HARD COPY AGAINST WEB SITE IMMEDL4TELY PRIOR TO EACH USE

VERWV HARD COPY AGAINST \VEB SITE IMMEDIATELY PRIOR TO EACH USE NSD 703 Nuclear Policy Manual Volume 2

b. Appendix Mmay be added to and reformatted as necessary to make it more applicable to a specific group. However, information shall not be removed from this appendix.

c. Document retention requirements do not apply to Appendix M.

d. The Reviewer s signature on the Procedure Change Process Record indicates that the procedure has been reviewed. in accordance with Appendix M.

5. An Applicability Deterrnniation (NSD 228) is NOT required for a minor procedure change,

6. The Renewer shall perform a detailed line by line review of all information changed as follows:

a. Ensure information contained within the procedure change is accurate and complete and that sufficient documentation is required by the procedure change to ensure the intent of the procedure is met.

b. Ensure WARNINGS or CAUTIONS are used appropriately to minimize risk to personnel or equipment.

c. Ensure the step(s) affected by the change can be accomplished ui the sequence tvriten.

d. Ensure the step(s) affected by the change meets the current Technical Specifications, UFSAR. SLCs, NSDs, Site Directives, etc.

e. Ensure the step(s) affected by the change provides for smooth interaction between site groups and efficient utilization of site resouces.

7. A. review of a minor procedure change does not require an extensive review of material not changed. Review the remaining parts of the procedure NOT changed to verify the following:

a. No missing procedure steps or pares

b. No obvious formatting problems created by the revision:

Inappropriate page breaks

Cautions, Warnings or Notes not on sanie page with applicable step

Incorrect step numbering

c. Change was appropriately incorporated into ALL affected parts of the procedure.

S. The Reviewer shall determine the need for cross-disciplinaw or additional reviews based on the following:

a. Response of a system under direct control of another group will be altered.

b. Steps in a procedure may affect the use or operation. of equipment under another croup control.

s 1

c. Another group will be required to provide personnel to assist in the performance of a procedure.

d. In cases where specific disciplines or training is needed other than that of the Reviewer to ensure a complete technical review of the change.

e. Procedures which affect or involve the Site Emergency Plan shall be reviewed by the Site Emergency Planning Section. Procedures winch mvolve Environmental Emergency Response plans (part of the Site Emergency Plan) shall also be reviewed by the Site Environmental Management Section.

26 REVISION 29 VERITY HARD COPY AGAINST WEB SITE IM.MEDLVIELY PRIOR TO EACH USE

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2010 MNS SRO NRC Examination QUESTION 98 KA_desc

( GEN2.3 Radiation Control Knowledge of radiological safety principles pertaining to licensed operator duties. such as containment

ent requirements. fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR: 41.12/

2.3.12 45.9/45.10)

SLC 16.11.20 (Gas Storage Tanks) limits the quantity of radioactivity in each Waste Gas Decay Tank (WGDT).

The basis for this limit assures the amount of radioactivity released would be substantially lower than the dose guideline values of Which One (1) of the following completes the statement above?

A. 10 CFR 20 during routine WGDT releases.

B. 10 CFR 100 during routine WGDT releases.

C. 10 CFR 20 in the event of a WG System leak or failure.

D. 10 CFR 100 in the event ofaWG System leak orfailure.

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2010 MNS SRO NRC Examination QUESTION 98 259 eneraI Discussion This SLC considers postulated radioactive releases due to a waste gas system leak or failure, and limits the quantity of radioactivity in each pressurized gas storage tank in the WASTE GAS HOLDUP SYSTEM to assure that a release would be substantially below the dose guideline values of 10 CFR Part 100 for a postulated event.

Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE: Answer is plausible because the normal release limits are delineated in 10CFR2O which provides the limits for what can routinely be released to the environment. It would reasonable for the applicant to misinterpret the basis of the limits for the quantity of radioactivity allowed to be stored in a WGDT to be contained in this document.

Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE: First part of the answer is correct, 10CFR100 is the basis for the dose guidelines. Second part of the answer is plausible because it would be reasonable for the applicant to associate the limits in SLC 16.11.20 to be associated with routine releases.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE: Answer is plausible because the normal release limits are delineated in 10CFR2O which provides the limits for what can routinely be released to the environment. It would reasonable for the applicant to misinterpret the basis of the limits for the quantity of radioactivity allowed to be stored in a WGDT to be contained in this document.

Second part of the answer is correct.

Answer D Discussion CORRECT: See explanation above.

Basis for meeting the KA

(/A is matched because the knowledge contained in the basis for this SLC requires the applicant to recall information that is directly related to ie radiological safety principle of the protection of the public by limiting the potential release of radioactive gases which could affect the public. Knowledge of the controlling document (10CFR100) and the limiting condition is a condition of a SRO license required to operate the station.

Basis for Hi Cog Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions (Rev 1 dated 03/11/2010) under the Screening Criteria for question linked to IOCFR55.43(b)(2) (Tech Specs):

1) It can NOT be answered solely by knowing < 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Tech Specs.

2) It can NOT be answered solely by knowing the LCO/TRM information listed above-the-line.

3) It can NOT be answered by knowing the Tech Spec Safety Limits or their bases.

4) It DOES require the applicant to have detailed knowledge of Tech Spec basis information to determine the correct answer.

Job Level Cognitive Level QuestionType Question Source SRO Memory BANK MNS Bank WEWGNO4 Development References Student References Provided OP-MC-WE-WG Obj. 6 SLC 16.11.20 KA KA_desc GEN2.3 Radiation ControlElKnowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (CFR: 41.12 /

2.3.12 45.9/45.10)

Friday, May 14, 2010 Page 268 of 275

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2010 MNS SRO NRC Examination QUESTION 98 259

-401-9 Comments: RemarkslStatus Friday, May 14, 2010 Page 269 of 275

Question 98

References:

From OP-MC-WE-WG Objectives OBJECTIVES OBJECTIVE 1 State the purpose of the Waste Gas (WG) System. WEWGOO1 X X X X X 2 Describe the system flowpath during normal operation, X X X X X shutdown operation and waste gas discharge. WEWGOO2 3 List four components that discharge waste gas into the WG X X X X X Header. WEWGOO3 4 List two types of non-radioactive waste gas discharged into X X X X X the WG Header. WEWGOO4 5 List the WG Discharge Flow Controller (WG-160) trips. X X X X X WEWGOO5 6 Concerning the Selected Licensee Commitments (SLC) related to the WG System:

. Discuss any commitments and their applicability. x x x

. For any commitments that have action required within one hour, state the action. x x x

. Given a set of parameter values or system conditions, determine if any commitment is (are) not met and any action(s) required within one hour.

. Discuss the basis for a given commitment.

SROonly x

WEWGOO7

16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.20 Gas Storage Tanks COMMITMENT The quantity of radioactivity contained in each gas storage tank shall be limited <49,000 Curies noble gases (considered as Xe-133).

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Quantity of radioactive A.1 Suspend all additions of Immediately material in tank not radioactive material to the within limit, tank.

AND A.2 Reduce the tank contents 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to within limit.

TESTING_REQUIREMENTS TEST FREQUENCY TR 16.1 1.20.1 Verify the quantity of radioactive material contained in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months each gas storage tank is within limit when radioactive materials are being added to the tank.

BASES This SLC considers postulated radioactive releases due to a waste gas system leak or failure, and limits the quantity of radioactivity in each pressurized gas storage tank in the WASTE GAS HOLDUP SYSTEM to assure that a release would be substantially below the dose guideline values of 10 CFR Part 100 for a postulated event.

Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tanks contents, the resulting total body exposure to a MEMBER OF THE PUBLIC at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Standard Review Plan 11.3, Branch Technical Position ETSB 11-5, Postulated Radioactive Releases Due to a Waste Gas System Leak or Failure, in NUREG-0800, July 1981.

REFERENCES None

Parent Question WEWGNO4 1 Pt SLC 16.11.20 limits the quantity of radioactivity in each Waste Gas Decay Tank (WGDT). What is the basis for this limit?

A. Assures the amount of radioactivity released would be substantially lower than the dose guideline values of 10 CFR 20 during routine WGDT releases.

B. Assures the amount of radioactivity released would be substantially lower than the dose guideline values of 10 CFR 100 during routine WGDT releases.

C. Assures the amount of radioactivity released would be substantially lower than the dose guideline values of 10 CFR 20 in the event of a WG System leak or failure.

D. Assures the amount of radioactivity released would be substantially lower than the dose guideline values of 10 CFR 100 in the event of a WG System leak or failure.

Answer 114 D

SLC 16.11.20 SRO only

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2010 MNS SRO NRC Examination QUESTION j 9 2599 KA_desc GEN2 3 Radiation Control L Iowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency or activities. (CFR: 41.12 I 43.4 I 45.10) 2.3.14 Given the following conditions:

A General Emergency has been declared

All Emergency Response facilities are activated

A non-licensed operator must be dispatched from the Operations Support Center to an area with an identified radiation field of 110 REM/hr in order to isolate the pathway for a large release to the environment

The operator will be in the area for approximately 15 minutes

All listed operators have volunteered and are fully aware of the nsks involved Which one of the following is the preferred operator to perform the task in accordance with RP/0/A/5700/004 (General Emergency) Enclosure 4.4 (Request for Emergency Exposure)?

A. Operator A is 26 years old and once in the past received a single acute dose of 26 REM TEDE.

B. Operator B is 52 years old and once in the past received a single acute dose of 27 REM TEDE.

C. Operator C is 29 years old and has a chronic lifetime dose of 18 REM TEDE.

D. Operator D is 49 years old and has a chronic lifetime dose of 19 REM TEDE.

Friday, May 14, 2010 Page 270 of 275

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2010 MNS SRO NRC Examination QUESTION 99 2599t

..eneral Discussion of 45 and those who normally encounter little exposure.

For this paicular case, Operator D, although younger than Operator B, has received the lowest cumulative dose and additionally has never vedan acute dose (which Operator B has).

Answer A Discussion INCORRECT. See explanation above.

PLAUSIBLE: This answer is plausible if the applicant concludes that younger person is more physically fit and thus more capable than an older person of receiving an acute dose without significant physical damage. An acute dose should be a once in a lifetime event (per Radiation Protection Policy Manual, Policy 11-02, Planned Special Exposure Dose Limits for Occupationally Exposed Personnel) and this person is not suitable with others available.

Answer B Discussion CT. See explanation abov e

PLAUSIBLE: This answer is plausible if the applicant concludes that the oldest person should be selected. However, an acute dose should be a once in a lifetime event (per Radiation Protection Policy Manual, Policy 11-02, Planned Special Exposure Dose Limits for Occupationally ed Personnel) and this person is not suitable with others availab le.

Answer C Discussion PLAUSIBLE: This answer is plausible if the applicant concludes that younger person is more physically fit and thus more capable than an older person of receiving an acute dose without significant physical damage, this would be correct since this individual has the lowest cumulative dose.

Answer D Discussion CORRECT. See explanation above.

ihe difference bewteen 18 and 19 R cumulative lifetime dose is negligible and the older candidate should be chosen.

Basis for meeting the KA The KA is matched because the applicant must have knowledge of the radiation protection hazards during an emergency condition and the eriafo_selection of the Operator who is most appropriate to receive an acute radiation_exposure.

Basis for Hi Cog This is a higher cognitive level question because it requires multiple mental steps. It requires the applicant to first recall from memory the actue exposure limits and requirements from RP-004 and then compare each answer to the recalled memory and to all of the other answers to select the most appropriate individual to receive the acute exposure.

Basis for SRO only This question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions (Rev 1 dated 03/11/2010) under the Screening Criteria for question linked to IOCFR55.43(b)(5) (Assessment and Selection of Procedures):

I) The question can NOT be answered by knowing systems knowledge.

2) The question can NOT be answered by knowing immediate Operator actions.

3) The question can NOT be answered by knowing entry conditions for any procedure.

4) The question can NOT be answered by knowing the purpose, overall sequence of events, or overall mitigative strategy of any procedure.

5) The question requires the applicant to have knowledge of the content of RP-004 General Emergency. Since this procedure would only be used by an SRO, the content of the procedure can NOT be considered RO level knowledge.

erefore, this is an SRO level question.

Job Level Cognitive Level QuestionType Question Source OComprehenANKQ98(Ban k259

)evelopment References Student References Provided

References:

1) RP/0/A!5700/004 General Emergency L__

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2010 MNS SRO NRC Examination QUESTION 99 25__

[KA KA_desc GEN2.3 Radiation ControhKnowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency or activities. (CFR: 41.12/43.4/45.10) 2.3.14 Comments: RemarkslStatus Friday, May 14, 2010 Page 272 of 275

Question 99

References:

From RP-004, Site Area Emergency:

£mdooreJ.4 C34 Reqiust br Ezmergemcv Exposare (a) Page 1 f)

Totil Effafr:oDoe Lez of Eye Other Oranr b:

Acthtv Egii,iea TEDEl All pa3r]) 10 rain 30 rein C(l ran.

operty LSe srata or 25 ram rain :ft sta,:dOL of Lrae PopuLKLoc Life satgor 5jrs *250nm Prcdeccin. eft.arEa Pupihthns (<

<: Eic:d dMcL&idprpnitizma

t: Jaattrrit.

<c; OV oa i voIut37 panc fl) owu; of tha ri,1 zxoht& AL b.cta bia

qpL ;j;r: ..tuown acn t sc. off i md trs,; mtho raoxxaLv Dro1m,tr itzI RP Badn No. arne Epl.o.r Strata of

.sditithril

.Iv :cium ia crr v a:kowL4cc nIbr;bicitfccdtharItnyb& xpa4&:o thu. :hcJn4átcL iabc. .d[yHfcJoa cSmk cobs 3rccmpih4ana oa.hDri-.k. afIi ;xpoua

,ckaowkdgs thu p]znod g;ocy (RPM ox da;:.. d.irur ox not; offrsti 1zthccizlboa) DnTin I. nçro:s th. ç.azn&d Ec!4ncrhpouze t (Emarnury Coota:oi or OY Dscrcz i attn ox nnc oft*a tin+/-a:i Dat; Tit ub çti Xadiaton ?rotsrtco Acton

- Diwmi; ztd £07 mk thcat atab.ancu

- tdat; rporü rq.irsaann psr 1CCF2O

to Sthdthiat fxpoun N:ton &ls

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2010 MNS SRO NRC Examination QUESTION 100__r, KA_desc GEN2.4 Emergency Procedures/PlanLlKnowledge of SRO responsibilities in emergency plan implementation. (CFR:

41.10 / 43.5 /

45.11) 2.4.40 Given the following plant conditions:

An Unusual Event was declared on Unit 2.

Initial Notification to the States, Counties and the NRC has been completed.

The Emergency Coordinator has just made the decision to upgrade the classification to an Alert The NRC is required to be notified immediately but no more than (1) after change of classification.

After the initial notification of the change in classification is made to the State and Counties, follow up notifications are required to be made every (2) until the emergency is terminated.

Which ONE (1) of the following completes the statements above?

A. 1. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

2. hour B. 1. ihour

2. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months C. 1. l5minutes

2. hour D. 1. 15 minutes

2. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Friday, May 14, 2010 Page 273 of 275

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 100 eneral Discussion n the scenario given, the applicant is presented with a situation where an NOUE was declared and all required initial notifications to the State, Counties, and the NRC has been completed. Subsequently, an escalation to an Alert occurs and the applicant is asked to evaluate the current notification requirements both to the NRC and the affect on the requirement for follow up notifications. Per our procedure, RP129 the follow-up notification requirement will change from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (NOUE) to a new requirement of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for the new Alert classification. The NRC notification procedure, RPI1O requires that the NRC be notified immediately but not more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after a change in classification.

Answer A Discussion CORRECT: See explanation above Answer B Discussion INCORRECT: See explanation above.

PLAUSIBLE: Part 1 is correct and therefore plausible.

Part 2 is plausible because the follow notification requirement for a NOUE is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months which was in effect prior to the upgrade in classification.

Answer C Discussion INCORRECT: See explanation above.

PLAUSIBLE: Part 1 is plausible because this is correct for every offsite agency except for the NRC. The applicant may misapply the 15 minute requirement to the NRC.

Part 2 is correct and therefore plausible.

Answer D Discussion IINCORRECT: See explanation above.

PLAUSIBLE: Part 1 is plausible because this is correct for every offsite agency except for the NRC. The applicant may misapply the 15 minute requirement to the NRC.

2 is_plausible_because the_follow_notification_requirement_for a NOUE is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months which was in effect prior to the upgrade in classification.

basis for meeting the KA The KA is matched because the applicant must have knowledge of the SRO (OSM) responsibilities for implementing the Emergency Plan (i.e.

fication requirements to offsite agencies after an esculation in emergency classification).

Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step. The applicant must first evaluate all of the information provided and then apply multiple rules to a change in a given situation. This requires the applicant relate understanding the rules pertaining to offsite notification and apply them to a dynamic situation.

Basis for SRO only This question is not tied to 10CFR5O.43 (b) but can be classified as an SRO Plant Specific Example. This question requires additional knowledge required for the higher license level and is unique to the SRO/OSM position. At MNS it is the responsability of the SRO to complete the notifications to offsite agencies and NRC notification to the NRC in the event that an emergency is declared. Per Lesson plan OP- MC-EP EMP (Emergency Plan) the objectives, #12 (Complete the ENF) and #14 (Complete the NRC event notification worksheet) are SRO ONLY objectives. (LPSO). Both the understanding of the requirements and the actual completion of the required paperwork along with the transmittal are SRO ONLY tasks at MNS.

Job Level , Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References itudent References Provided iinbjective:

l)EP-EMP#l2 and#l4 29 (Notifications to Offsite Agencies From The C/R)

.c.4.2Pg1of8 RP!lO (NRC Immediate Notification Requirements)

[nc4.1 Pg 1 of 14 Friday, May 14, 2010 Page 274 of 275

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2010 MNS SRO NRC Examination QUESTION 100__

KA KA_desc Emergency Procedures/PlanDKnowledge of SRO responsibilities in emergency plan implementation. (CFR:

4LlO/

Zints:

1 i RemarkslStatus 0

Friday, May 14, 2010 Page 275 of 275

Question 100

References:

OP-MC-EP-EMP Obj: 12 & 14 OBJECTIVES NNLLL OBJECTIVE L L P P 0 OORSR ROO io Given the EPIPs and the emergency situation, classify the x x event.

ii Given the EPIPs and the emergency situation, provide the X X appropriate Protective Action Recommendations (PARs).

12 Given the EPIPs and the emergency situation, complete the X X Emergency Notification Form.

13 Describe the use of the Selective Signaling Phone System to X X X X X notify the State and County.

14 Given the EPIPs and the emergency situation, complete the X X appropriate portions of the procedure for an NRC Event Notification Worksheet.

15 State the requirements for Initial and Follow-up Notifications X X X including:

. Time Requirements

. Agencies to be contacted

From OPIO!B157001029 Enc. 4.2 Pg lof 8 Enclosnre 4.2 RRO/W5700/029 Completion and Transmission of a Page 1 of S Tollow-up Message NOTE: New initial messages for higher classification upgrades are addressed in Enclosure 4.1. {PlP M-Ol-3711}

1. Make follow-up notifications according to the following table:

Follow-up Notifications

1. Follow-up notifications to the State(s) and Counties must be made according to the following schedule:

-For a NOUE. every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months until the emergency is terminated. For ALERT. SAE or GE every hour until the emergency is terminated.

OR

-If there is any significant change to the situation (make notification as soon as possible).

OR

-As agreed upon with an Emergency Management official from each individual agency Documentation shall be maintained for any agreed upon schedule change. The interval for ALERT. SALE, and GE shall not be greater than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to any agency.

2. If a follow-up is due and an upgrade to a higher classification is declared, there is no need to complete the follow-up ENF. In this case, the offsite agencies must be notified that the pending follow-up is being superseded by an upgrade to a higher classification and information will be provided.

3. Follow-up messages in the General Emergency classification that involve an upgrade in PARs must be communicated to the offsite agencies as soon as possible and within 15 minutes.

I Complete an Emergency Notification Form by one of the following:

El Obtain a preprinted ENE.

OR C Obtain a blank ENF.

From OP!01B157001010 Enc. 4.1 Pg lof 14 Enclosure 4.1 RpOAf700iolo Events Requiring NRC Notification Page 1 of 14 tTLi Events Requitino L\NEDL4TENOTIHCATIONS: REPORTING ThE REQUIREMEYfS REPORTABLE EVENTS Correspondinr 10CFRSectoninlirackets []

4.1.1.1 The declaration of any oftheEmergencv Classes 1.1.1.1 lmndiatelyafterootifirationto state(s) andlocal goveoiament (counties) spec+/-eMcGuire Emercencv Plan and not later than one hour after the time the EmezgencvClassyas

[51.72a(l*)J declare& Immediately report inc change from one Emergency Class to mother or a termination of the Emergency Class. (Use Enclosure 4.2.)

or See follow upreqtiirements in section 4.1.6 of this procedure.

2c(1;(ii)j 7

[51. anychange from one EnaergencvClass to another or 150.72c(lxiii)] a termination of the Emergency Class 4,1.1.2 Eventsinvolangreceiviaag and openangpackages 4.1.1.2 NOTE: Reporting wader IOCFR2O.1906 slanddhema& as follon the containing quantities ofradioactive material in excess of licensee shall immediately notify the final delivery camer and by 111.19061 a Type A quantity as defined in section 71.4 and telephone and telemanc mailgrank or facsimile and the NRC Operations appendix A to part 71 of this chapter when Center at 9-301-516-5100.

[20.190(1)) 1) Removable radioactive surface contamination exceeds the limits of section 7I.S7Q) of 10CFRIO:

or

[10.1906d(2)] 2) External radiation levels exceed the limits of section 71.47 of this chapter.

4.1.13 Anylost stolen. ormissinglicensedmatetialin on 4.1.1.3 Immediarelyafterits occurrence becomes knownto thelicensee.

aggrerate quantity equal to or greater than 1.000 limes 110.2201a(i the quantity specifledin appendixCto part 10 under such circumstance that it appears to the licensee that an eosure cruldresnktopons intmrestiactedaoeas.

01

[20.220la(ia)J 3) Within 30 days after the occanrence afanylost. stoiei or missinc licensed material becomes known to the licensee. all licensed material ma quantity greater than lltianesthequantityspecifleduaappendixCtopaztll that is still missinc at this time.

ML102650109

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