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Initial Exam 2010-301 Final Ro/Sro Written Exam Working Copy
	ML102640737
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
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I-1 N AL. ccy (f? tJtL-kt i5 CCfi.y FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 1 i1 ,=SYSOO3 K5.O1 - Reactor Coolant Pump System (RCPS) iowledge of the operational implications of the following concepts as they apply to the RCPS: (CFR: 41.5 /45.7) ne relationship between the RCPS flow rate and the nuclear core operating parameters (quadrant power tilt, imbalance, DNB rate, local power density. difference in loop T-hot pressure) Given the following conditions on Unit 1:

The unit is operating at 40% RTP
NCP C trips on overcurrent Assuming no operator action, which ONE (1) of the following describes the effect on the Departure from Nucleate Boiling Ratio (DNBR) AND reactorthermal power?

A. DNBR will INCREASE. Reactor power decreases and stabilizes at a new lower thermal power. B. DNBR will DECREASE. Reactor power decreases and stabilizes at a new lower thermal power. C. DNBR will INCREASE. Reactor power initially decreases and then returns to 40% thermal power. D. DNBR will DECREASE. Reactor power initially decreases and then returns to 40% thermal power. Tuesday, July 13, 2010 Page 1 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 1 General Discussion The decrease in Reactor Coolant Flow with reactor power, temperature (core delta-T), and pressure remaining the same will cause a decrease in DNBR. In this case Actual Heat Flux (AHF) remains the same while the Critical Heat Flux (CHF)(amount of heat required to cause a departure from nucleate boiling) will decrease. Therefore DNBR (CHF/AHF) decreases. Since steam demand has not changed core thermal power (Q=mcpdelta-T) must remain the same steady-state to steady-state. However, reactor power will initially decrease due to the immediate effect of the loss of flow (mass flow rate decreases) while core delta-T initially has not changed. After the initial decrease in reactor thermal power, the colder water returning to the reactor will cause an increase in reactor power, core delta-T will increase, and core thermal power will return to 40% thermal power based on steam demand. The increase in core delta-T will result in the water at the core exit being closer to vaporization and therefore CHF decreases causes an additional

decrease in DNBR.

The conclusion is that DNBR decreases and reactor power will initially decrease and then return to 40% thermal power. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: The first part is plausible If the applicant confuses the DNBR with an actual departure from nucleate boiling as the likelihood of an actual departure from nucleate boiling has increased. Part 2 is plausible if the applicant neglects to consider the long-term effect of the NC pump trip. The reactor power will initially decrease due to the decrease in flow. However, power will not stabilize at the new lower power but will return to 40% thermal power since steam demand has remained constant. Answer B Discussion rCOR1ECT: See explanation above. PLAUSIBLE: First part is correct. DNBR does decrease. Part 2 is plausible if the applicant neglects to consider the long-term effect of the NC pump trip. The reactor power will initially decrease due to the decrease in flow. However, power will not stabilize at the new lower power but will return to 40% thermal power since steam demand has remained constant. . Answer C Discussion - INCORRECT: See explanation above. PLAUSIBLE: The first part is plausible If the applicant confuses the DNBR with an actual departure from nucleate boiling as the likelihood of an actual departure from nucleate boiling has increased. The second part of answer is correct. Answer D Discussion CORRECT: See explanation above. -___________________________________________________________ Basis for meeting the KA . - The KA is matched because the applicant must determine the effect of a reduction in NC system flow rate (due to NC pump trip) on core operating parameters (i.e. DNBR & core thermal power). Basis for Hi Cog -- -- This is a higher cognitive level question because it requires more than one mental step to arrive at the correct answer. The applicant must first determine the effect of the reduction in NC system flow on the DNBR. Then the applicant must determine the long-term effect on reactor thermal power. Basis for SRO only - - Job Level , Cognitive Level QuestionType - Question Source RO Comprehension NEW Tuesday, July 13, 2010 Page 2 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 1 2501 velopment References Student References Provided

.earning Objectives: OP-BNT TH08024 & Th08026

References:

1. OP-BNT-THO8 Section 3.1

2. OP-BNT-RTO8 Section 2.3

3. OP-BNT-THO7 Section 5.0 SYSOO3 K5.01 - Reactor Coolant Pump System (RCPS)

Knowledge of the operational implications of the following concepts as they apply to the RCPS: (CFR: 41.5 /45.7) The relationship between the RCPS flow rate and the nuclear core operating parameters (quadrant power tilt, imbalance. DNB rate, local power density, difference in loop T-hot pressure) 401-9 Comments: RemarkslStatus L -9 Comments: No comment. Resolution / Comments: N/A - Tuesday, July 13, 2010 Page 3 of 294

Question I

References:

From OP-BNT-THO8 Section 3.1: The exact value for CHF is somewhat nebulous as many design factors affect CHF. Equations have been developed which allow CHF to be calculated for both PWR (subcooled coolant) and BWR (saturated coolant) reactor designs. Many of the variables which affect the magnitude of CHF are design factors beyond the scope of this lesson. However, there are four parameters which are controlled by the operator. These parameters are coolant pressure, temperature, flow and power. Pressure: saturation temperature and pressure are related. As pressure increases, the saturation temperature also increases. Increasing pressure affects the density of both water and vapor and therefore, the formation of vapor bubbles on the heat transfer surface. Pressure also affects the subcooling margin (see note below) and the time required for a vapor bubble to form or collapse. Within the normal range of operating pressures, increasing pressure increases the value of CHF. Bulk coolant temperature: bulk coolant temperature affects CHF in much the same manner as pressure. As the temperature of the coolant increases, less additional heat energy is required to cause vapor formation, so the rate of vapor formation increases, increasing the likelihood that the clad surface will develop a vapor film. As bulk coolant temperature increases, CHF decreases. Mass flow rate: as mass flow rate increases, the velocity of the coolant increases. The increased velocity increases turbulence and reduces the boundary layer thickness which in turn reduces the temperature gradient in the boundary layer. The increased velocity also reduces the size of the vapor bubbles which form on the surface of the fuel clad by sweeping them off the surface and supplies subcooled liquid to the fuel surface. The net effect is increased heat transfer efficiency as the difference in temperature between the fuel surface and the coolant in the boundary layer is maximized. CHF increases as coolant flow rate increases. Power: power affects both the Actual Heat Flux (AHF) of the fuel and CHF. As power increases the rate of heat production increases as does AHF. As AHF increases the fuel clad surface temperature increases as does the rate of vapor bubble formation on the clad surface. The increased rate of vapor formation on the fuel clad surface increases the probability of a vapor film forming on the heat transfer surface (DNB). For a given coolant velocity and temperature, CHF decreases as power (heat flux) increases due to the increase in clad surface temperature. At higher power, the coolant temperature rise across the core increases. Regardless of what method of reactor coolant temperature control is used (ramped or constant), the temperature and enthalpy at the core exit is higher than the inlet. As the coolant passes through the core gaining heat, CHF steadily decreases from the bottom of the core to the top of the core.

From Lesson Plan OP-BNT-RTO8 Section 2.3: Once the reactor has operated at power, xenon and other fission product poisons begin to build to their equilibrium values. This buildup adds negative reactivity to the reactor. Boron concentration changes or control rod withdrawal must compensate for this negative reactivity. The control rods are normally kept almost fully withdrawn to maintain the axial flux difference (axial imbalance or AT at ONS) within limits to optimize fuel utilization. Xenon buildup is normally compensated for by dilution. Since xenon buildup is a relatively slow process, this presents no significant problem for the reactor operator.

                                     *Rarnp SIG LevelsTave Constant               )-

T Ramp 602°

REACTOR POWER 100% Figure 2 Reactor Coolant Temperature verses Reactor Power (B&W) 2.3 LOAD CHANGES AT POWER Objectives 4, 5, Once power passes the point of adding heat into the power range of operation, the combined effects of moderator temperature feedback and fuel temperature feedback cause the reactor characteristics to change. Now, secondary system steam demand controls the steady state reactor

power level. For power to be stable (constant), two conditions must be satisfied: 1) the reactivity must be balanced (that is, keff 1) and 2) the power mismatch must be 0. The power mismatch (PMM) is defined to be the difference between heat production by the reactor and heat removal by steam demand. PMM = QPRoD QREM Equation 1

From Lesson Plan OP-BNT-THO7 Section 5.0: BLOVOWN BLOWDOWN FLOW FLOW (M) () (M Figure 7 Reactor Heat Balance Objective 17 Two methods are commonly used to calculate thermal power based on the available parameters. These methods use the heat transfer equations developed earlier. Reactor coolant system parameters can be used to determine CTP, using the following equation: Q =thcAT Where, for a reactor coolant system heat balance: Q = heat transfer rate of the reactor core (Btu/hr)

         +/-  =  mass flow rate of the reactor coolant system (lbm/hr) cp  =  specific heat capacity of the reactor coolant (Btu/lbm-°F)

AT = temperature difference across the core (TH T) (°F) Equation 16 Plant instrumentation provide m and AT values. The heat capacity of the water, c can be closely approximated using steam tables and exact values are available in subcooled water tables. The heat transfer rate across the steam generator can also be used to determine CTP. Since boiling occurs on the secondary side of the steam generators, the calculation must account for the latent heat addition as well as the sensible heat addition which occurs as the subcooled feedwater is first heated to boiling, then vaporized. For this reason, the change in enthalpy between the steam and the feedwater is normally used to perform the calculations

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 2 25O2 SYSOO4 K6.02 Chemical and Volume Control System

nowledge of the effect of a loss or malfunction on the following CVCS components: (CFR: 41.7/45.7) jernineralizers and ion exchangers Given the following conditions on Unit 1:

Unit is operating at 100% RTP
The controller for I KC-1 32 (Letdown Hx Outlet Temp Ctrl) has been placed in MANUAL due to erratic operation
Subsequently, NV letdown flow is increased by 10 GPM as requested by Chemistry As letdown temperature increases, NC system boron concentration will (1) AND if letdown temperature continues to increase, letdown flow will automatically bypass the demineralizer at (2)

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

2. 120°F

,      B.          1. INCREASE

( 2. 138°F C. 1. DECREASE

2. 120°F D. 1. DECREASE

2. 138°F Tuesday, July 13, 2010 Page 4 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 2 L 25O2 neraI Discussion Requires operator to determine the effect of increasing letdown temp on the MB Demineralizer. At low temperatures, the boron affinity is increased. At high temperatures, boron affinity is reduced. If the temperature is increased previously captured boron ions are released from the MB Demineralizer thus increasing NC system boron concentration. If Letdown Hx outlet temperature increases to 138°F, 1 NV-i 27A will divert to the VCT to protect the demineralizer resin. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Answer 1 is correct. Answer 2 is plausible because the Letdown Hx Outlet Hi Temperature annunciator (1AD-7 / H2) alarms at 120°F. Answer B Discussion CORRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: Answer 1 is plausible if applicant does not understand the effects of increasing letdown temperature on the demineralizer resins affinity for boron atoms. It is plausible for the applicant to conclude that increasing temperature (which would cause the demineralizer resin to expand) would result in a larger surface area in the resin and thus increase the probability of boron absorption. Answer 2is plausible because the Letdown Hx Outlet Hi Temperature annunciator (1AD-7 / H2) alarms at 120°F. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: Answer 1 is plausible if applicant does not understand the effects of increasing letdown temperature on the demineralizer resins affinity for boron atoms. It is plausible for the applicant to conclude that increasing temperature (which would cause the demineralizer resin to pand) would result in a larger surface area in the resin and thus increase the probability of boron absorption. Answer 2 is correct. Basis for meeting the KA The KA is matched because a malfunction has occurred (temperature controller failure) and the applicant must determine how the malfunction affects the CVCS components in question (in this case the demineralizers). The applicant must also recall that letdown will be diverted to the VCT on high temperaturetoprotectthedemineralizerresin. __________________________ Basis for Hi Cog LBasis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory MODIFIED 2009 MNS RO Retake [eveloPment References Student References Provided

References:

1. OP-MC-PS-NV Section 2.6 Learning Objective 4 2. OP-BNT-CHO5 Section 7.3 Learning Objective CH05025 SYSOO4 K6.02 Chemical and Volume Control System

Knowledge of the effect of a loss or malfunction on the following CVCS components: (CFR: 41.7/45.7) Demineralizers and ion exchangers 1-9 Comments: - RemarkslStatus 401-9 Comments: No comment. Tuesday, July 13, 2010 Page 5 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 2 2502, Resolution / Comments: N/A Tuesday, July 13, 2010 Page 6 of 294

Question 2

References:

From Lesson Plan OP-MC-PS-NV: 2.6 Letdown Heat Exchanger I Objective # The letdown heat exchanger cools the letdown flow to the operating temperature of the mixed bed demineralizers. Letdown flow is through the tube side of the heat exchanger and KG (Component Cooling) flows through the shell side. The temperature sensor controls KG-I 32 and regulates the amount of cooling so that IO5°F is maintained. The temperature setpoint is inserted using the DCS Graphic - NV Charging Control KC-i 32 L/D Hx Outlet Temp Control Graphic or its SLIMs station.

2.7 NV-I 24 Letdown Pressure Control Valve The letdown pressure control valve, NV-124, reduces letdown pressure to within design limits. During normal operation, it maintains a backpressure of approximately 350 psig. This valve is used in conjunction with NV-459 when initiating or adjusting letdown flow. Pressure can be controlled manually or automatically. During fixed letdown orifice operation, backpressure is increased to 450 psig to minimize vibration The valve has proportional trim which increases pressure control response time at low flows because changes in valve position at the lower end of the valve stroke results in only small changes in the flow coefficient. As the valve, NV-124, becomes more open, a change in valve position results in a much larger change in flow coefficient which makes the valve more responsive to changes in letdown flow. (PIP-0-M96-0861) NV-i 56, the relief downstream of NV-i 24, protects piping and demineralizers from overpressure. It has a setpoint of 255 psig and relieves to the VCT. 2.8 NV-I 27A (Demin Temp. Divert Valve) NV-i 27A auto diverts NV flow to the VCT if high temperature (i38°F) exists in the letdown line to prevent damage to demineralizers resin. 2.9 Reactor Coolant Filters A and B Objective # 4 There are two (2) filters located in the letdown line. Only one is placed in service at a time (normally NC A). The filters collect resin fines and particulates to prevent resin from reaching the NV pump suction. There are two (2) local AP indicators provided. While both filters are currently utilized as post-filters (downstream of the demineralizers), the B filter can be aligned as a pre-filter to the demineralizers. Demineralizer bypass lines are provided to allow continued letdown filtration with the demineralizers out of service.

From Lesson Plan OP-BNT-CHO5 Section 7.3: boron saturated resin, the ability of the resin to exchange unwanted impurities is severely reduced. One of the primary reasons for this demineralizer is to control the trace amounts of chloride (also an anion) present in the reactor coolant. When a chloride ion enters a boron saturated resin bed only some of the chloride ions will be exchanged due to the large number of borate ions present which compete for the exchange sites on the resin. The amount of chloride which can be retained is dependent on the concentration of the competing ion, borate. For this reason, early in core life when the concentration of boron in the reactor coolant is high, the demineralizer is not able to remove all the chloride from the reactor coolant. As the core ages and the concentration of boron is reduced, the concentration of chloride in the reactor coolant decreases as the amount of competing borate ions decreases. Boron affinity of a resin bed is also affected by the temperature of the coolant as it passes through the bed. At lower temperatures, the borate ion bonding to exchange site contains three boron atoms. At higher temperatures, the borate ion contains only one boron atom. The results of this characteristic are that at lower temperatures, resins are more efficient at removing boron from coolant than at higher temperatures. A boron saturated resin bed will actually release boron as the temperature is increased. The second chemical added to the reactor coolant is lithium hydroxide. Lithium is a cation with a single + charge. During normal operation the cation portion of the mixed bed demineralizer is saturated with lithium. In a similar fashion to the borate / chloride competition, the relatively large amount of lithium present in the reactor coolant reduces the capacity of the mixed bed demineralizer for other unwanted cations which are present in only trace amounts (such as cesium). One reason for the chemical and volume control system cation bed demineralizer is to remove cesium as the unit is shutdown for refueling (to prevent radioactive cesium isotopes from presenting dose problems to workers). The cation bed demineralizer is NOT lithium saturated and can effectively remove lithium, cesium, and other trace cation impurities from the coolant. In the discussion of the Effects of Ion Exchange, the effect of passing a sodium chloride solution through various types of resin beds was discussed. Based on that discussion it follows that a demineralizer can be used to alter the pH of the process fluid. This is commonly done as a means to control the pH of the reactor coolant. The reactor coolant is a solution of boric acid with lithium hydroxide added to increase the pH. Lithium hydroxide is produced in the coolant via a boron neutron reaction. This production of lithium causes

the pH of the reactor coolant to increase. One way to reduce the concentration of lithium (and the pH) in the reactor coolant is to process the coolant through a hydrogen form cation demineralizer. The lithium ions are removed and replaced with hydrogen ions (which then form water), effectively reducing the pH. In systems where it is possible to subject the demineralizer resin to high temperatures, demineralizers have automatic features to protect against temperature damage. This is usually accomplished by automatic closure of the demineralizer inlet valves to isolate the demineralizer from high temperature liquid when high temperature at the inlet to the demineralizer is sensed. These systems are typically equipped with bypass valves that can divert flow around the demineralizer until normal system temperature is restored.

bjective 26 Oconee has reactor coolant demineralizers which are loaded with only anion resin. These demineralizers are used primarily to remove boron from the reactor coolant late in the core cycle. The amount of feed and bleed of the reactor coolant necessary to lower the boron concentration

Question 2 Parent Questions: Question 2243 (2009 NRC RO Retake Exam): FOR REVIEW ONLY DO NOT DISTRIBUTE - 2009 MNS RO NRC Retake Exarnina QUESTION 43 2243 KA KA_desc APEO26 Ability to determine and inteipret the following as theyapplyto thcLoss of Component Cooling Waler: (CFR: 43.5 45.13):Thc nonnalvalnes and uppcr limits for the Tclnpcianwcs of the components ccsDledby CCW AA2.04 Given the following current conditions on Un 1:

Unit is operating at 100% RTP
A malfunction of the Letdown Hx Outlet temperature controller has caused IKC-132 (Letdown Hx Cooling Water Control) valve to slowly drift closed
Letdown Heat Exchanger Ouet temperature has increased from 106°F to 115°F Which ONE (1) of the following correctly completes the statement below?

Based on current conditions, NC system temperature will (1) due to reactivfty effects AND it Letdown Hx Outlet temperature continues to increase, 1NV-127A, LD Hx Outlet 3-Way Cntrl will divert to the VCT at (2) A. (1) decrease (2) 120°F B. (1) decrease (2) 138°F C. (1) increase (2) 120°F D. (1) increase (2) 138°F TIiursdav March 25 2010 Page 98 of 171

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2009 MNS RO NRC Retake Examina QUESTION 43 I 22451 General Discussion The increase in Letdosvtt He at Exchanger Outlet temperature causes an increase in mixed bed dentineaahzer resin temperature. This temperature increase results in thertnal regeneration of the resin and tire release of boron from the detnineralizer resin to the letdown. This results in au increase in the boron concentratron of the chargina water going back to the NC system which causes NC system temperature to decrease. If Letdown temperature increases to LISt. letdown will drvert to the VCT to protect the dennueralizer resin frona damage. Tins K.A is matched because the applicant insist detennitte the effect that increase in letdown line temperature will base ott NC systena temperature anti the tippet lhnit for letdown temperar.ae before NV-l2A divens This is a comprehension level questton because the applicant mttst process anti evaluate multiple pieces of infotination to cletenntne die correct anssver. Etrst. the applicant naust detenaune the increase m letdown tettiperatore svtll result tn a release of boron front the demitterahzers and then cletennitte that the increase in boron concentration in the NV cltaring will result in a temperature decrease. The applicant must then recall front memory the t eanperature setpomtat for the chversiort of letdown tiow. Answer A Discussion htconect NC system reanperature decreasing is correct. The temperatsue of 120 t as plausible because that us thc seiponir for the Letdown Hz Onlet Hi Teanpeaatatre Aimunciator. Answer S Discussion CORRECT, Answer C Discussion lateotreet. Plaustlale if die applicant does not recall the effect of letdown lute teniperatuare on tIme affinity of denttneralizer resna for boron. I2OCF as plausible hecatise it us the setporiit for the Letdown Hz Otulet Tenaperanne Hi Annunctator. Answer D Discussion Incorrect. Plausible if the applicant does not recall the effect of letdown line ten,eramre on the anity of deiuineralizer resin for boron. roe temperature setpoittt is correct. Basis for meeting the KA Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension XEW Development References Student References Provided Anminciaaor Response Procedute for Panel tAD-7 H2 BNT-CHOSR5. Ion Exchange Objective 23 page 23 KA KAdesc APEO26 Ability to detenuine and iaaterpret the fohlosvnmg as they apply to the Loss of Component Cooling Water: (CFR: 43.5 45.l3):The normal valttes and wpper limits for the tentpeiatnres of the conapotuents cooled by CCW AA2.04 401-9 Comments: RemarkslStatus APEO26A.O4 Disazree with Lead on this one. facility feels that 120 deg has I thinlc l2OoF is NOT plausible because it is too low. Cotisider using a ample plausibility, plait to discuss Earthier silani datrtng Ott site higher temperature. At what temp does tlae resin bunt? Sotnetlutig higher reviesv. tItan 138 svonld also be acceptable. REA 10.28.09 Thursday, Match 25, 2010 Page 99 of 171

Question 14 (2009 NRC Exam): Examination Outline Cross-reference: Level RO SRO x x Tier# 2 DRAFT Group 4* 1 K/A# 004K5.50 Importance Rating 2.6 Chemical and Volume Control System: Knowledge of the operational implications of the following concepts as they apply to the NV system: Design basis letdown system temperatures: resin integrity Proposed Question: Common 14 1 Pt Given the following on Unit 1:

1KC-132 (L/D HX Cooling Water Control Valve) has failed
LD HX outlet temperature is 115°F and increasing If LD Hx outlet temperature reaches (1) 1NV-127A (LD Hx Outlet 3-Way Temp Cntrl) will AUTO divert letdown flow to the (2)

A. (1) 138°F (2) VCT to protect the demineralizer resin B. (1) 138°F (2) RHT to protect the VCT from over temperature C. (1) 120°F (2) VCT to protect the demineralizer resin D. (1) 120°F (2) RHT to protect the VCT from over temperature

Proposed Answer: A Explanation (Optional): As letdown temperature increases to 138°F 1NV-127A will divert letdown flow to the VCT to protect the demineralizer resin. A. Correct. B. Incorrect: See explanation above. Plausible because there is a diversion flow path to the RHT. However, that diversion of letdown flow is on VCT high level. C. Incorrect: See explanation above. Plausible because 1NV-127A does divert to the VCT on high temperature. Also, the Letdown Heat Exchanger Outlet Hi Temperature annunciator alarms at 120°F. D. Incorrect: See explanation above. Plausible because there is a diversion flow path to the RHT. However, that diversion of letdown flow is on VCT high level. Also, the Letdown Heat Exchanger Outlet Hi Temperature annunciator alarms at 120°F. Technical Reference(s) LP OP-MC-PS-NV (Rev 55) Pg 35 of (Attach if not previously provided) 153 Section 2.14 OP/1/A/6100/O1O H (Annunciator (Including version or revision 1*) Response for Panel 1AD-7) Pg 37 of 52 (Rev 57) Proposed references to be provided to applicants during examination: None Learning Objective: OP-MC-PS-NV Obj 10 (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 41.5 55.43 Comments:

Chemical and Volume Control System: Knowledge of the operational implications of the following concepts as they apply to the NV system: Design basis letdown system temperatures: resin integrity KA is matched because the candidate must be familiar with the letdown design basis maximum temperature. He/she must also know the operational implication (letdown auto divert to the VCT) should that temperature be reached.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 3 SYSOO5 A 1.02 Residual Heat Removal System (RHRS)

bility to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls cluding: (CFR: 41.5 / 45.5) RHR flow rate Given the following conditions on Unit 1:

Unit is in MODE 5 with ND Train A in service
1 NI-i 73A (Train A ND to A & B CL) is OPEN
ND flow is being reduced by throttling i ND-29A (A ND HX Outlet) in preparation for removing ND from service 1ND-68A (A ND Pump & A HX Mini-flow) will open if 1A ND pump flow decreases to less than a MAXIMUM of (1) . The Operators in the Control Room can verify that 1ND-68A has opened by recirc flow indication on (2)

Which ONE (1) of the following completes the statements above? A. 1. 325 GPM

2. the OAC ONLY B. 1. 325GPM

2. a chart recorder on MC-7 AND the OAC C. 1. 750 GPM

2. the OAC ONLY D. 1. 750GPM

2. a chart recorder on MC-7 AND the OAC Tuesday, July 13, 2010 Page 7 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 3 2503 General Discussion To assure that no damage to the pump will occur due to overheating or vibration during low flow operation (less than 7SOgpm), a 3 inch mini-flow line is provided from the outlet of the each ND HX back to the associated pump suction. The flow through this line is automatically controlled by ND-68A (A ND Pump & A HX Mini-flow) for Train A and ND-67B (B ND Pump & B HX Mini-flow) for Train B. These valves will open when its associated pump flow decreases to 750 gpm and will close when flow increases to 1400 gpm. This mini-flow line works well provided the check valve at the discharge of the ND pump does not become closed due to a higher pressure downstream. This can occur when both ND pumps are running with their discharge lines cross connected ( such as ND 15 and ND3O open ) and their pressure/flow characteristics are significantly different. The stronger pump will force the other pumps discharge check valve closed thus its 3 inch mini-flow line will be disabled. To account for this, a 2 inch mini-flow line around each ND pump is provided. This line is always in operation to provide a minimum of 300 gpm but targeted for 325 gpm flow (set by a manual throttle valve) back to the suction of its respective pump. This flow path is upstream of the check valve therefore can not be isolated by it. Indication for both the 3 inch ND PUMP and HX Mini-flow (0 to 500 gpm) and 2 inch ND Pump Mini-flow (0 to 500 gpm) lines is provided on MC 7 in the Control Room via chart recorders NDCR5O6O for pump A and NDCR5O7O for pump B. Recirc flow indications are also provided on the OAC. -_________________________________ Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part I is plausible because the is the design flow rate through the mini-flow line around the ND pump only. Part 2 is plausible if the applicant does not recall that there is also recirc flow indication on MC-7. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because this is the design flow rate through the mini-flow line around the ND pump only. trt 2 is correct. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: Part I is correct. Part 2 is plausible if the applicant does not recall that there is also recirc flow indication on MC-7. Answer D Discussion CORRECT: See explanation above. Basis for meeting the KA The KA is matched because RHR flow rate is being changed and the applicant demonstrates the ability to monitor changes in parameters by knowing when the mini-flow recirc valve should open to protect the pump from damage (i.e. exceeding design limits) and knowing what indications are available to determine if the mini-flow valves have operated correctly. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided srning Objectives: PS-ND #4,5

References:

1) Lesson Plan OP-MC-PS-ND Section 2.1 Tuesday, July 13, 2010 Page 8 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 3 _.SYSOO5 Al.02 Residual Heat Removal System (RHRS) 7 - j ility to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Ri-IRS controls Juding: (CFR: 41.5 / 45.5) RI-IR flow rate {iOl-9 Comments: emarksIStatus L 401-9 Comments: No comment. Resolution / Comments: N/A Tuesday, July 13, 2010 Page 9 of 294

Question 3

References:

From Lesson Plan OP-MC-PS-ND Section 2.1: pump breaker must be closed before the A(B) ND PUMP LOW FLO TO COLD LEGS alarm on AD9 (setpoint 500 gpm ) can occur. This alarm also looks at the position of valves ND-15B, 30A (Train B(A) ND to Hot Leg Isol ) and NI-I 73A ( Train A ND to A & B CL), 178B (Train B ND to C&D CL ) to actuate the alarm. The logic diagram for this alarm is located in the Alarm Response Manual on its annunciator window data sheet. To assure that no damage to the pump will occur due to overheating or vibration during low flow operation (less than 75Ogpm), a 3 inch mini-flow line is provided from the outlet of the each ND HX back to the associated pump suction ( refer to Drawing 7.1 ). The flow through this line is automatically controlled by ND-68A (A ND Pump & A HX Mini-flow) for Train A and ND-67B (B ND Pump & B HX Mini-flow) for Train B. These valves will open when its associated pump flow decreases to 750 gpm and will close when flow increases to 1400 gpm. The mini-flow line loop includes the ND HX to ensure the recirculating fluid does not become overheated due to the energy added by the pump. This mini-flow line works well provided the check valve at the discharge of the ND pump does not become closed due to a higher pressure downstream. This can occur when both ND pumps are running with their discharge lines cross connected (such as ND15 and ND3O open) and their pressure/flow characteristics are significantly different. The stronger pump will force the other pumps discharge check valve closed thus its 3 inch mini-flow line will be disabled. To account for this, a 2 inch mini-flow line around each ND pump is provided. This line is always in operation to provide a minimum of 300 gpm but targeted for 325 gpm flow (set by a manual throttle valve) back to the suction of its respective pump. This flow path is upstream of the check valve therefore can not be isolated by it. Indication for both the 3 inch ND PUMP and HX Mini-flow (0 to 500 gpm) and 2 inch ND Pump Mini-flow (0 to 500 gpm) lines is provided on MC 7 in the Control Room via chart recorders NDCR5O6O for pump A and NDCR5O7O for pump B. Recirc flow indications are also provided on the OAC. Both ND pumps will auto start on a Safety Injection Signal. This automatic start signal is received from the Diesel Generator Sequencer. If power is lost to the train related 4160V buss ( Blackout), the ND pumps will receive a start permissive from the sequencer so they can be manually started if needed, but do NOT auto-start. 2.2 ND Heat Exchangers There are two heat exchangers per unit (one per train). Each heat exchanger is designed to remove one-half of the total heat load. They are shell and U-tube type heat exchangers with ND flowing through the tube side and KC through the shell side. An annunciator LO KC FLOW TO A ND HEAT EXCH and LO KC FLOW TO B ND

HEAT EXCH on AD9 warns the operator if KC flow to the ND heat Exchanger decreases to 4500 gpm. This alarm is state sensitive (it will not generate an alarm under conditions when an alarm would not be applicable) and has logic to determine if ND flow exists on that train and will only alarm if KC flow is needed to that train.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 4 SYSOO5 K3.O1 Residual Heat Removal System (RHRS)

nowiedgeof the effect tha loss malfunction of the RHRS will have on the following: (CFR: 41.7 /45.6) Given the following conditions on Unit 1:

The unit is in MODE 4
The crew is increasing NC system temp and pressure for unit startup
ND Train A is in service
NC system temperature is being maintained at 140°F If instrument air is lost to 1 ND-34 (A & B ND Hx Byp) the valve will fail (1) AND NCsystemtemperaturewill (2)

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

2. INCREASE B. 1. CLOSED

2. INCREASE C. 1. OPEN

2. DECREASE D. 1. CLOSED

2. DECREASE Tuesday, July 13, 2010 Page 10 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 4

..eneral Discussion ND-34 fails open on a loss of instrument air. This bypasses flow around the ND Hx causing a decrease in flow through the heat exchanger. This causes the temperature of the water returning to the NC system to increase and therefore NC temperature would increase.

Answer A Discussion CORRECT: See explanation above. Answer B Discussion ___________________ INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible If the applicant does not understand how ND-34 fails on a loss of air. It is plausible for the applicant to conclude that ND-34 fails closed on a loss of VI since this would prevent the ND Hx from being robbed of flow. Part 2 is plausible if the applicant does not recall the flowpath for the bypass line. For example, if the bypass line tapped off downstream of the heat exchanger and diverted back to the pump suction (common design for many Westinghouse RHR system Hx bypass lines and the same as the arrangement for the ND pump recirc valves), it would decrease the flow through the Hx if the bypass valve was initially throttled and then failed closed. Therefore, it is plausible for the applicant to conclude that NC system temperature_would increase in this case. Answer C Discussion rnbORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because ND-34 does fail open on a loss of instrument air. Part 2 is plausible if the applicant does not understand the flowpath for the Hx bypass line. For example, if the bypass line tapped off downstream of the heat exchanger and diverted back to the pump suction (common design for many Westinghouse RHR system Hx bypass lines and the same as the arrangement for the ND pump recirc valves), it would increase the flow through the Hx if the bypass valve was initially throttled or closed and then failed open. Therefore, it is plausible for the applicant to conclude that NC system temperature would decrease in this case. Answer D Discussion CORRECT: See explanation above. PLAUSIBLE: Part I is plausible If the applicant does not understand how ND-34 fails on a loss of air. It is plausible for the applicant to conclude that ND-34 fails closed on a loss of VI since this would prevent the ND Hx from being robbed of flow. IfND-34 closed on a loss of VI, flow would increase through the ND Hx causing NC system temperature to decrease making Part 2 plausible. Basis for meeting the KA is matched because a failure of an RHR system component has occurred (ND-34 failing open) and the applicant must determine the effect that this malfunction has on NC system temperature. Basis for Hi Cog This is an analysis level question because it requires more than one mental step. The applicant must first recall from memory how ND-34 fails on a loss of air. The applicant must then determine from that failure how ND system flowrate through the ND Hx is affected and the resultant effect on NC system temeperature. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided OP-MC-PS-ND Section 2.3.6 Learning Objective PSNDOO8 1 SYSOO5 K3.Ol Residual Heat Removal System (RHRS)

owledge of the effect that a loss or malfunction of the RHRS will have on the following: (CFR: 41.7 / 45.6) CS Tuesday, July 13, 2010 Page 11 of 294

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2010 MNS SRO NRC Examination QUESTION 4 401-9 Comments: Remarks!Status 14019 Comments: No comment. Resolution / Comments: Tuesday, July 13, 2010 Page 12 of 294

Question 4

References:

From Lesson Plan OP-MC-PS-ND Section 2.3.6: ND-I 8 and ND-33 are used during residual heat removal mode of operation to control bypass flow around ND Heat Exchanger B and A respectively. Opening ND-18 and ND-33 would allow the respective trains ND heat exchanger to be bypassed during the ECCS recirculation mode if a loss of instrument air were to occur ( since bypass valve ND-34 fails open upon a loss of instrument air). Therefore, these valves are required to remain closed during Modes 1 3, when the ECCS system is required. If opened during

Mode 4 for residual heat removal temperature control, they shall be capable of manual closing upon ECCS actuation. If opened for residual heat removal mode, these valves shall be closed prior to swapover to sump recirculation mode of ECCS operation, for the respective ND train to be operable. Valve status is also provided to the OAC. 2.3.6 ND-34 (A & B ND HX Bypass) This valve can be operated from MCII or the ASP by a manual loader. This valve is used in conjunction with ND-14 and ND-29 to control NCS cooldown rate and temperature. ND-34 will fail open on a loss of Instrument Air (VI). ND-34 is regulated to maintain a constant return flow to the NCS. A constant flow rate allows the ND pumps to continuously operate on a more efficient part of their performance curve. Flow through this return line is higher during the initial stages of NCS cooldown to limit the ND System heatup rate, and thus thermal shock to the ND heat exchangers. This valve is not required for the unit to achieve cooldown and is therefore not safety related. 2.3.7 ND-15B (Train B ND to Hot Leg Isol), ND-30A (Train A ND to Hot Leg Isol) These motor operated valves are controlled from the ND section of MCI I in the Control Room by open/close pushbuttons. These fail as is valves provide cross tie isolations for the ND Trains. These valves have no auto open/close control features. These valves are opened in standby readiness, but closed in cold leg recirc. On an ECCS actuation, the ND System must be capable of providing flow to all four NCS loops (even with single failure). By having ND-15B and ND-30A open, either ND pump is capable of supplying all four NCS loops. Therefore, closing either ND-I5B or ND-30A in Mode 1, 2, or 3 will make both ND trains inoperable. An alarm is actuated on the BOP panel whenever either of these valves reaches the closed position. 2.3.8 ND-67B ( B ND Pump & B HX Mini-flow) and ND-68A (A ND Pump & A HX Mini-flow) These safety related, normally closed motor operated valves are interlocked to automatically open on a train related pump start when ND flow through its train related ND heat exchanger falls below the 750 gpm setpoint ( as sensed by NDFT525O for pump A and NDFT526O for pump B). When flow reaches the 1400 gpm setpoint or if the associated pump stops, the valve will close.

2.3.9 ND-35 (ND System to FWST lsoation) This valve is an 8 manually operated gate valve. ND-35 is used during outage periods to transfer water from the reactor coolant system or refueling canal to the refueling water storage tank. ND-35 is also used as a gravity flow path from the

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 5 SYSOO6 Kl.14 Emergency Core Cooling System (ECCS)

nowledge of the physical connections and/or cause-effect relationships between the ECCS and the following systems: (CFR: 41.2 to 41.9/ p5.7 to 45.8) lAS Concerning the operation of Engineering Safeguards Modulating Control Valves: Upon receipt of a (1) signal, the modulating control valve circuit will (2) the control valves. Which ONE (1) of the following completes the statement above? A. 1. Safety Injection ONLY

2. maintain VI aligned to B. 1. Safety Injection ONLY

2. ventairoff C. 1. Safety Injection OR Blackout

2. maintain VI aligned to D. 1. Safety Injection OR Blackout

2. vent air off Tuesday, July 13, 2010 Page 13 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 5 5O5j General Discussion The modulating control valve circuitry controls the solenoids supplying air to selected safety related control valves. These control valves are normally controlled by other non-safety controllers and instrumentation. In order to prevent these non-safety controls from causing the control valves to assume an improper position after a safety event, these safety related solenoids valves will vent air off its control valve to cause it to assume its safe position These solenoid valves de energize upon receipt of a safety injection signal from the DIG load sequencer Answer A Discussion INCORRECT: See explanation above PLAUSIBLE: First part is correct and therefore plausible. Second part is plausible because some of the modulating valves contain individual VI accumulator tanks to ensure the valves do not reposition in the event of a loss of VI. It would therefore be reasonable for the applicant to believe that VI would be aligned to the affected valves. Answer B Discussion CORRECT: See explanation above Answer C Discussion INCORRECT: See explanation above PLAUSIBLE: First part includes a B!O signal in addition to the SI signal. Both are ESF signals, both come from the sequencer and both result is the repositioning of many safety related valves. It would be reasonable for the applicant to believe the modulating control valves are affected by both signals. Second part is plausible because some of the modulating valves contain individual VI accumulator tanks to ensure the valves do not reposition in the event of a loss of VI. It would therefore be reasonable for the applicant to believe that VI would be aligned to the affected valves. Answer D Discussion INCORRECT: See explanation above AUSIBLE: First part includes a B/O signal in addition to the SI signal. Both are ESF signals, both come from the sequencer and both result is the repositioning of many safety related valves. It would be reasonable for the applicant to believe the modulating control valves are affected by both signals. Second part of the answer is correct and therefore plausible. Basis for meeting the KA The K/A is matched because the applicant is being tested on the cause-effect relationship between an ECCS actuation signal and the resulting effect on ECCS valves which are supplied by the lAS. The signal results in a change in alignment of the VI supply to these control valves. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided OP-MC-PSS-RN Page 39 (Rev 43) From OP-MC-PSS-RN Page 41 (Rev 43) OP-MC-PSS-RN Obj: 12 SYSOO6 Kl .14 Emergency Core Cooling System (ECCS)

owledge of the physical connections and/or cause-effect relationships between the ECCS and the following systems: (CFR: 41.2 to 41.9 / j.7 to 45.8) lAS Tuesday, July 13, 2010 Page 14 of 294

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2010 MNS SRO NRC Examination QUESTION 5 401-9 Comments: - RemarkslStatus Comments: Distractors A and C: Change align VI to Maintain its alignment because some of the modulating valves do not reposition in the event of a loss of VI. This will lend the 2nd part of distractors A and C more plausible. Resolution I Comments: Changed align VI to to maintain VI aligned to. See attached for proposed_revision. .___________ Tuesday, July 13, 2010 Page 15 of 294

Question 5

References:

OP-MC-PSS-RN Obj: 12 OBJECTIVES NNLLL OBJECTIVE L L P P 0 OORSR ROO 8 Describe the RN System Flow path ( suction source, essential and non-essential header alignment and discharge point) for the following:

        . Normal operation                                         X X X X
        . Operation following a Blackout                           X X X X X
        . Operation following a Safety Injection                   X X X X X 9  Explain the reason for taking a suction on the low level intake. X X X X 10  Concerning the RN essential and non-essential headers:
       . List the loads supplied by each header                   x x x x
       . Identify which loads are automatically supplied on a     x x x x x Blackout, Safety injection and/or Phase B.

11 Explain the reason for NOT isolating the auxiliary building X X X X X non-essential header on a Blackout signal. 12 Describe the operation including any interlocks for the X X X X X following valves:

       . RN42A ( AB Non Ess Supply Isol)
       . RNI71B (B D/G Supply Isol)
       . 1RN1 ( Low Level Intake Isolation)
       . Engineering Safeguards Modulating Control Valves and Reset Circuitry 13  Describe the operational concerns when cycling RN valves             X X X that are shared between Unit 1 and Unit 2.

14 Given a parameter associated with the RN system, describe X X X X the indications for that parameter. 15 Given a Limit and Precaution associated with the RN System, X X X X X discuss its basis and when it applies.

From OP-MC-PSS-RN Page 39 (Rev 43) The KC HX Supply isolation Valves (RN-86A, 1 87B) have an AUTO/MANUAL mode select switch and an open/close pushbutton on MCi 1. The open/close pushbuttons are only operable when the mode switch is in the Manual position. If the mode select switch is in the AUTO position, the valve will auto open when the train related RN pump starts and will receive a signal to close when the train related RN pump is stopped. In either the AUTO or MANUAL mode of operation, these valves will automatically open upon receipt of a Blackout or Safety injection signal. Also, the Blackout and Safety injection signal is interlocked with the AUTO portion of the valve closure circuitry to prevent the valve from automatically closing while a Blackout or Safety injection signal is still present. The valves are normally selected to the AUTO mode. Low Level intake isolation valve 1 RN1 is a non-safety related MOV controlled from I MCi 1 by a pushbutton which is covered to prevent operation of the valve except in an emergency. The valve is wired through two breakers in MCC SMXL. The breaker in compartment 3C is normally disconnected which allows power to the valve to be disconnected while still leaving control power available for position indication in the Control Room. Therefore in order to close I RN 1, power must be restored by reconnecting the breaker in MCC SMXL compartment 3C and using the manual close pushbutton. If maintenance activities require shifting RN suction to the RC cross over or SNSWP such that 1RN1 will be closed, compensatory action will be required for Train A to prevent specific valves from automatically re-aligning to LLI on a Ss or BO signal. ORN-4AC (Train lB & 2B RC Supply) and ORN-148AC (Train 1A & 2A Disch to RC) will automatically open on SSF transfer. In addition to ORN-4AC and ORN148AC, valves ORN-147AC, ORN-283AC, ORN-3O1AC, URN 1OAC, and URN 12AC can be operated from the SSF. The RN controls and indications located in the Standby Shutdown Facility will be covered in Lesson plan OP-MC-CP-AD. The Train A(B) Engineering Safeguards Modulating Control Valve Reset Pushbuttons and reset lights are located on the RN section of MCi 1. The modulating control valve circuitry controls the solenoids supplying air to selected safety related control valves. These control valves are normally controlled by other non-safety controllers and instrumentation. In order to prevent these non-safety controls from causing the control valves to assume an improper position after a safety event, these safety related solenoids valves will vent air off its control valve to cause it to assume its safe position. These solenoid valves de-energize upon receipt of a safety injection signal from the D/G load sequencer. The modulating reset circuitry has a mechanical latching relay which will maintain the valves in their safe position after the safety injection signal is reset. To gain control of these valves, the safety injection signal must be reset and the operator must depress the train related modulating valve reset pushbutton. The indicating light is labeled RESET and is normally illuminated. Upon receipt of a Safety Injection Signal, the light will be off. Following reset of the latching relay, the light will illuminate. Failure of the fuse in the pushbutton circuit renders all modulating valves inoperable. PIP O-M96-2018 in section 5.2 covers an operating experience associated with these fuses.

From OP-MC-PSS-RN Page 41 (Rev 43) The following are the Train A modulating valves: Safe Position

RN-89A (RN to A KC HX Control) Open*
RN-22A (RN Strainer A Backflush Automatic Drain Isol) Close**
ND-.29 (A ND HX Outlet) Open
KC-57A (A ND HX Return) Open The following are the Train B modulating valves: Safe Position
RN-i 90B (RN to B KC HX Control) Open*
RN-26B (RN Strainer B Backflush Automatic Drain lsol) Close**
ND-14 (B ND HX Outlet) Open
KC-82B (B ND HX Return) Open
Theses valves open to their travel stop position.

**

Position of these valves does not affect backwash capability. 2.6.2 Cycling Shared Valves I Objective#13 I When performing the cycling of shared RN valves the following items need to be considered to prevent undesired system alignments.

Ensure RN is aligned per the unit specific operating procedure to allow valve cycling.
Ensure that an adequate flow path exists for operating components. Review OAC graphics and any other pertinent information to evaluate the effects of the valve stroke. Do not rely on the VST or functional test to accomplish this task.
Ensure the opposite units RO evaluates the cycling of shared RN valves prior to cycling the valve.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 6 SYSOO6 K6.02 Emergency Core Cooling System (ECCS)

nowledge of the effect of a loss or malfunction on the following will have on the ECCS: (CFR: 41.7/45.7) Pore flood tanks (accumulators) Unit 1 is operating at 100% RTP. Given the following indications for Unit 1 CLAs: IA lB 1C ID Pressure 630 PSIG 570 PSIG 590 PSIG 615 PSIG Level 7305 GAL 6970 GAL 6890 GAL 7375 GAL Which ONE (1) of the following describes how the ECCS system is affected (if at all) by the CLA parameters listed above? A. 1 B CLA ONLY is INOPERABLE. B. 1 C CLA ONLY is INOPERABLE. C. 1A and 1C CLAs are INOPERABLE. D. 1 B and 1 D CLAs are INOPERABLE. Tuesday, July 13, 2010 Page 16 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 6 2506 neral Discussion JAW Tech Spec 3.5.1, Accumulators, the required range of level for a Cold Leg Accumulator to be OPERABLE is greater than or equal to 6870 gals and less than or equal to 7342 gal. Based on level indications, CLA ID is INOPERABLE. The required pressure range for a Cold Leg Accumulator to be OPERABLE is greater than or equal to 585 PSIG and less than or equal to 639 PSIG. Based on the pressure indications, CLA 1W is INOPERABLE. Answer A Discussion - INCORRECT: See explanation above PLAUSIBLE: This answer is plausible if the applicant does not recall the pressure and level bands for the CLAs. lB CLA being INOPERABLE is correct. However, 1D CLA is also INOPERABLE because its level is high out of the Tech Spec required band. If the applicant does not recall the level and pressure bands correctly, they can conclude that ID CLA level is within limits. Answer B Discussion -____________ INCORRECT: See explanation above PLAUSIBLE: This answer is plausible if the applicant does not recall the pressure and level bands for the CLAs. If they do not recall the bands correctly, they can conclude that IC CLA is INOPERABLE since its level is the lowest of all CLAs. Answer C Discussion INCORRECT: See explanation above PLAUSIBLE: If the applicant does not recall the pressure and level ranges for CLA operability, this answer is plausible since CLA IA has the highest pressure and CLA iC has the lowest level. If the applicant does not know the correct ranges, they could conclude that CLA iA pressure was out-of-spec high and that CLA 1 C level was out-of-spec_low. However, both pressure and level for CLA IA and IC are in_spec. Answer 0 Discussion CORRECT: See explanation above. Basis for meeting the KA - e K/A is asking for the effect of a loss or malfunction on the Core Flood Tanks (Accumulators) on the ECCS. Since the CLAs are part of the ( CS as defined by UFSAR Section 6.3.2.2.1) anything that affects the operability of the CLAs in turn affects the operability of the ECCS and hence its ability to perform its design function. Thus, CLAs with operating parameters outside of their Tech Spec limits affects the ability of ECCS to perform its design function. Therefore, the KA is matched. -. Basis for Hi Cog This is a higher cognitive level question (JAW NUREG- 1021 Appendix A. Step 3 .C.c) because the applicant must associate multiple data points. First, the applicant must recall a setpoint from memory (in fact two ranges of setpoints). Then, the applicant must compare the data given in the stem of the question to the recalled setpoints to arrive at the correct answer. Since the question requires two mental steps to answer the question correctly, this is a higher cognitive level question. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO . Comprehension . NEW Development References Student References Provided Tech Spec 3.5.1, Accumulators SYSOO6 K6.02 Emergency Core Cooling System (ECCS)

Knowledge of the effect of a loss or malfunction on the following will have on the ECCS: (CFR: 41.7 / 45.7) Core flood tanks (accumulators) Comments: Remarks/Status 0l-9 Comments: Change A to YB ONLY is inoperable Change B to 1C ONLY is inoperable All or None are poor distractors and rarely picked if applicant is Tuesday, July 13, 2010 Page 17 of 294

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2010 MNS SRO NRC Examination QUESTION 6 not sure. lB and IC have the lowest pressure and level respectively. Delete (if at all) from the stem This Q is U because there are 2 NP distractors Resolution / Comments: Changed question per Lead Examiners recommendation. See attached file for proposed revision which includes new distracter analysis for A and B. Tuesday, July 13, 2010 Page 18 of 294

Accumulators 3.5.1 Question 6

References:

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Accumulators LCO 3.5.1 Four ECCS accumulators shall be OPERABLE. APPLICABILITY: MODES 1 and 2, MODE 3 with RCS pressure> 1000 psig. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One accumulator A.1 Restore boron 72 hours inoperable due to boron concentration to within concentration not within limits. Ii m its. B. One accumulator B.1 Restore accumulator to 24 hours inoperable for reasons OPERABLE status. other than Condition A. C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time of Condition A or B AND not met. C.2 ReduceRCSpressureto 12 hours

                                        < 1000 psig.

D. Two or more D.1 Enter LCO 3.0.3. Immediately accumulators inoperable. McGuire Units 1 and 2 3.5.1-1 Amendment Nos. 218/200

Accumulators 3.5.1 SURVEILLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator isolation valve is fully open. 12 hours SR 3.5.1.2 Verify borated water volume in each accumulator is 12 hours

              > 6870 gallons and < 7342 gallons.

SR 3.5.1.3 Verify nitrogen cover pressure in each accumulator is 12 hours

              > 585 psig and < 639 psig.

SR 3.5.1.4 Verify boron concentration in each accumulator is within 31 days the limits specified in the COLR. AND NOTE Only required to be performed for affected accumulators Once within 6 hours after each solution volume increase of> 1% of tank volume that is not the result of addition from the refueling water storage tank SR 3.5.1.5 Verify power is removed from each accumulator isolation 31 days valve operator when RCS pressure is> 1000 psig. McGuire Units 1 and 2 3.5.1-2 Amendment Nos. 184/1 66

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 7 25O7 SYSOO7 K5.02 Pressurizer Relief TanklQuench Tank System (PRTS)

iowledge of the operational implications of the following concepts as they apply to PRTS: (CFR: 41.5 / 45.7) .lethod of forming a steam bubble in the PZR Given the following conditions on Unit 1:

A unit startup is in progress following refueling
The crew is preparing to draw a bubble in the Pressurizer.
NC system pressure is 360 PSIG
NC system is in Solid Ops with LTOP in service
The 1A NC pump is RUNNING

1. Per Selected Licensee Commitment 16.5-4 (Pressurizer), what is the MAXIMUM allowable Pressurizer heat up rate?

2. Based on current plant conditions, how are non-condensable gases removed from the NC system?

A. 1. 75°F in anyone hourperiod

2. Cycle Pressurizer PORV5 B. 1. 75°F in any one hour period

2. Cycle the Reactor Vessel Head vents C. 1. 100°F in anyone hour period

2. Cycle Pressurizer PORV5 D. 1. 100°F in any one hour period

2. Cycle the Reactor Vessel Head vents Tuesday, July 13, 2010 Page 19 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 7 2507 eneraI Discussion There is no tie between the pressurizer and the PRT when forming a bubble at McGuire. The Pzr is taken water solid first with non condensibiles removed via the PORVs, then heated to saturation while water solid. After SU-8 is entered with the Pzr water solid, non-condensable gasses are removed from the Reactor Coolant System via the Rx Head Vents. In SU-8 with NC pumps running, the PORVs would never be used to vent the NC system. Then letdown flow is increased to draw the bubble. PZR Heatup rate per SLC 16.5-4 is 100 degrees F in any 1 hour period. ________________________________ Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because 75°F in one hour is administrative limit for PZR heat up. Part 2 is plausible because cycling the PORVs would vent non- condensibiles but once SU-8 is entered and a NC Pump is placed in service, this is not an option. -_________________________ Answer B Discussion -__________________ INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because answer 75°F in one hour is administrative limit for PZR heat up. Part 2 is correct. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is correct. Part 2 is plausible because cycling the PORVs would vent non- condensibiles but once SU-8 is entered and a NC Pump is placed in service, this s not an option. nswer D Discussion [LORRECT. See explanation above. Basis for meeting the KA - KA is matched because the candidate is required to know the methodology for removing non-condensable gasses from the NC System prior to bubble formation. This along with testing knowledge of the PZR Heatup rates examines the operational implications of forming a steam bubble in the PZR. Again, at MNS there is no tie between the Pressurizer and PRT with specific regards to drawing a bubble. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided OP/2/A/6100/SU-8, Heatup to 200 degrees F, Rev. 30, Enclosure 4.2, page 3 Lesson Plan OP-MC-PS-NC, Reactor Coolant System, Rev. 30, page 19 OP-MC-PS-NC Obj. 4 SYSOO7 K5.02 Pressurizer Relief Tank/Quench Tank System (PRTS)

Knowledge of the operational implications of the following concepts as they apply to PRTS: (CFR: 41.5 /45.7) ethod of forming a steam bubble in the PZR 401-9 Comments: RemarkslStatus i9 Comments: Tuesday, July 13, 2010 Page 20 of 294

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2010 MNS SRO NRC Examination QUESTION__7 25O No comment. Resolution / Comments: N/A Tuesday, July 13, 2010 Page 21 of 294

Question 7

References:

NNLLL OBJECTIVE L L P P 0 OORSR R00 1 State the purpose of the Reactor Coolant System. X X X X 2 Describe the flow path in the Reactor Coolant System with all X X X NC Pumps running and with less than all pumps running. 3 Describe the indications which would be used to detect a X X X X reactor vessel head 0-ring leak and how this line can be isolated. 4 Concerning the manual and remote reactor vessel head vents:

       . state their purpose including when each would be used      X X X
       . state how the vents are operated                           X X X X 5  Sketch the Reactor Coolant System and include all                X X X X penetrations and instrumentation associated with system operation and control per Drawing 7.5.

6 State the purpose of the pressurizer. X X X X 7 Describe how the inherent characteristics of the pressurizer X X X X reduces the effects of pressure transients. 8 Explain why the surge line is connected to the NC hot leg and X X X X why the spray line is connected to the NC cold leg. 9 Statethepurposeofmaintainingaconstantsprayflowtothe X X X X pressurizer. 10 State how Pzr spray flow will be effected if only A or B NCP is X X X X operating. 11 State the purpose of providing the capability of auxiliary spray X X X X flow to the pressurizer.

From Lesson Plan OP-MC-PS-NC: 2.0 COMPONENT DESCRIPTION 2.1 Reactor Vessel The reactor vessel is cylindrical with a hemispherical bottom and a removal hemispherical flanged top. The vessel contains the core (fuel), core support structures, control rods, neutron shield pads and four (4) inlet and four (4) outlet nozzles. The flow through the core enters the inlet nozzles, down the core barrel-wall annulus and turns up at the bottom of the vessel and travels up the core and then out the outlet nozzles. The inlet and outlet nozzles are spaced 45 apart and are tapered to reduce loop pressure drop. The vessel design provides the smallest, most economical volume to house the internal components. I Objective#3 I The reactor vessel flanged top is sealed by two metallic 0-rings. To monitor the integrity of this seal, leak detection is provided by two leakoff connections. One connection samples between the inner and outer 0-ring and the other samples outside the outer 0-ring. Both of these lines combine to a common header which has a manual isolation valve, NC23, normally open during operation ( refer to Drawing 7.2). The leakoff line between the two seals has an isolation valve, NC-25A , which can be closed from 1(2)MC-l0 if leakage is excessive. The other leakoff line has a manual isolation valve, NC24, which is normally open. During normal operation, excessive seal leakage is detected by a temperature detector which will provide an alarm on 1 (2)AD6 Rx Vessel Flange Leakoff Hi Temp, if the line temperature increases 200 F above ambient. A meter on 1 (2)MC 10 provides indication of leakoff temperature. I Objective # The reactor vessel head has two vent methods, manual and remote (refer to Drawing 7.3). The local manual head vent is provided to ensure air is removed from the reactor vessel head area during NCS fill. This line has a flow sight glass to provide indication that the vessel is vented. The remote operated vent is comprised of two branches in parallel (train A & B ) with two solenoid valves in each branch. Both trains (valves NC 272A,C, 273A,C, 274B, and 275B) can be operated from the control room panel 1(2)MCO5. These valves are used to vent the reactor vessel head to PRT during accident conditions. The A train vent valves are used as a letdown path to control Pressurizer level during SSF operation. The reactor vessel and internals are covered in more detail in lesson plan OP-MC-PS-RVI.

Enclosure 4.2 Heatup to 200°F (Control Room Activities) Page 3 of 15 C 3.3 Continue NC System heatup to 110 - 195T by controlling ND System flow and EC flow to ND Hx. NOIE Non-condensible gases will continue to collect in Reactor Vessel Head for several hours. 3.4 To remove non-condensible gases from Reactor Vessel Head. perxodically perform the following: 3.4.1 Open the following:

2NC-272AX (Tm 2 k Head Vent 1 to PRT Isol)
2NC-273A.C (Tm 2A Head Vent to PRT Isol)
2NC-2743 (Tnt 23 Head Vent to PRY Isoli
2NC-2 753 (Tm 23 Head Vent to PRY Iso])

3.4.2 WHEN visible increase in PRY level observed without appreciable increase in PRT l>re55111e. close the following:

2NC-272A.C (Tm 2A Head Vent to PRI Isol)
2NC-2 73A.C (Tm 2A Head Vent to PRT Isol)
2NQ-2743 (Tm 23 Head Vent to PRY Iso])
2NC-2753 (Tnt 23 Head Vent to PRY Isol) 3.5 WHEN openino Reactor Head Vent Solenoid Valves no longer effective in removing non-condensible gases from Reactor Vessel Head normal operating conditions have been established in the PRY per OP 2A/6150004 (Pressurizer Relief Yank).

perform the following: 3.5.1 (lose 2NC-Sl (PRY Vent). 3.5.2 Remove temporary filter unil from 2NC51. 3.5.3 Install pipe cap on 2NC5 1. 3.6 Notify Primary Chemistiy to check Pzr Oxygen concentration less than 100 ppb. Person NotifIed Date Time Unit 2

Ellelosure 4.2 1 00/SD-S Heatup to 200OF (Control Room Activities) Page 4 of 15 3.7 IF Pzr Oxygen Concentration is greater than 100 ppb D additional hydrazine is needed in the Pzr. perform the following: 3.7.1 Open 2NV-21A (NV Spray to Pzr Isol). 3.7.2 Ensure the following closed: 2NV-13B (NV Supply to A NC Loop Isol) 2NV-1OA (NV Supply to D NC Loop Isol) 2NC-27 (A Loop Pzr Spray Control) 2NC-29 (B Loop Pit Spray Control) NOTE; The following step will ensure sufficient flow via 2V-2lA NV Spray to Pzr Isol) to acid hydrazine to the Pzr while protecting NC Pump seals. 3.7.3 Maintain a minimum of 6 gpiu NC Pump seal injection flow by adjusting the fo llowin.g:

2NV-241 (Seal Inj Flow Control
2NV-23S (Charging Line Flow Control) 3.7.4 Notify Primary Chemistry to add hycirazine.

Person Notified Date Tune 3.7.5 WHEN hyclrazine added. open:

2NVl3B (NV Supply to ANC Loop Isol) (odd cycle)

OR

2NV- IdA (NV Supply to D NC Loop Isol) (even cycle) 3.7.6 Close 2NV-21A (NV Spray to Pzr Isol).

3.7.7 Opeta spray valve that will provide niaxinmni flow to provide Pit mixing:

2NC-29 (B. Loop Pzr Spray Control)
2NC-27 (A Loop Pzr Spray Control,i
2NV-S4OA (ND to Pzr Anx Spray Control)

Unit 2

F]iclowre 4.2 o 2 A 6100 SU-2 Heatup to 200°F (Control Room Activities) Page 5 of 15 3.7.2 Maintain Pzr level and NC Pump seal injection flows by adjusting the following:

DNV-241 (Seal Inj Flow Control)
2NV-238 Charging Line Flow Control) 3.7.9 Notif Primary Chemistry to inform Control Room when Pzr Osvgen concentration less than 100 ppb.

Person Notified Date lime 3.8 Stop 2A Containment Aux Carbon Filter Fan. 3.9 Draw a bubble in the Pzr as follows: 3.9.1 Ensure PT 2 A 4600 005 (Surveillance Requirements For Unit Heatup) in progress. 19.2 Monitor the following parameters: C Letdown flow (M2A0764) C Charging flow (M2A0758) C Pzr Surge Line tenaperature (M2A0S55) C Pzr Steam Space temperature (M2A0849) C Pzr Water Space temperature (M2A0843) C Pzr Surge Line Pzr Water temp D T (M2P4322)

C WR NC System pressure (M2A0S26) C Low Range NC System pressure (M2A0845) C VCT level (M2A0734) 19.3 Ensure the following:

NC System Tavg 110 i9 5*-F and stable

NC System presstn-e 3 IC) 340 psig and stable
Letdown flow 80 100 gpnm

NC System Oxygen concentration less than 100 ppb
Pzr Oxyzen concentration less than 100 ppb 3.9.4 Adjust 2NV- 121 ND Letdown Controll to control NC Systent pressure 310 340 psig wlnle frilly opening 2NV-124 (Letdown Pressure Control).

C 19.5 Maintain NC System lnessnre 310 - 340 psig. Unit 2

7 16.5 REACTOR COOLANT SYSTEM 16.5.8 Pressurizer COMMITMENT The pressurizer temperature shall be limited to:

a. A maximum heatup of 100°F in any 1-hour period,

b. A maximum cooldown of 200°F in any 1-hour period, and

c. A maximum spray water temperature differential of 320°F.

APPLICABILITY At all times. REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. NOTE A.1 Restore pressurizer 30 minutes All Required Actions temperature to within limits. must be completed whenever this Condition AND is entered.

        ---                          A.2   Peorm engineering             72 hours Pressurizer temperature             evaluation to determine not within limits,                  effects of the out-of-limit condition on the structural integrity of the pressurizer.

AND A.3 Determine that the 72 hours pressurizer remains acceptable for continued operation. B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Reduce pressurizer 36 hours pressure to < 500 psig.

OP.iiA/6lO0!SU-S Page3 of 4 ileatup to 200 Degrees F

1. Purpose Direct activities to begjn heating The NC System to 200F.

2. Limits and Precautions.

2.1 This procedure is Reactivity Manasement related because it controls activities that can affect core reactivity by changing NC System temperature. (R.Mj 12 PD Pinup operation while in LTOP mode is prohibited unless directed by an El> or AP. (overpressurization concern) {PIP M95-0541 }) 23 WHEN one or more Pzr PORVs and associated isolation valves are open. heatup rare is limited to less than or equal to SOTbr (Administrative) and less than or equal to 6WEhr (Tech Spe4 14 With NC System temperature greater than 2001. KC flo.w to each ND Mx shall be greater than 2000 gpm. 15 Exceeding LOOT in the NC System until at least one NC Pump is in service is prohibited to minimize cold water addition to the Reactor Core resulting in positive reactivity addition and pressure excursion during water solid operation when starting an NC Pump. 2 Minimize Dir between S/Gs secondaiy inventozy and operating ND trains END to NC Cold Leg prior to starting NC Pumps. {PIP 99-5022} 17 Tn LOW PRESS Mode. PzrPORVs will open onNC narrow range pressure between 378 382 psig. Narrow range pressure can be monitored by CAC points M2A1359

(2NC-328) and M2A1365 (2NC-34A). 2.S MizaiimunVCl pressure is 15 psigvith NC: Pump(s) in service. 2.9 Maximum NCDT pressure is S psig with NC Pump(s) in service. 2.10 Mininuun NCDT and PRT pressure is 0 psig. (PIP 99-5074) 2.11 Pressurizer heatup rates shall be less than 75Thr (SLC limit is lOOTilu). 2.12 Maximum Boron concentration difference between NC System and Pzr is 50 ppm. 2.13 IF temperature difference between Pr and spray fluid is greater than 320T use of Auxiliary Spray is prohibited.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 8 SYSOO8 K4.02 Component Cooling Water System (CCWS)

nowledge of CCWS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) iperation of the surge tank, including the associated valves and controls Concerning the operation of 1KC-122 (KC Surge Tank Vent Valve): When a (1) alarm is received on 1 EMF-46A(B), 1 KC-122 will automatically close and the valve (2) Which ONE (1) of the following completes the statement above? A. 1. Trip 1

2. must be locally re-opened B. 1. Trip I

2. will automatically re-open when the alarm clears C. 1. Trip 2

2. must be locally re-opened D. 1. Trip2

2. will automatically re-open when the alarm clears Tuesday, July 13, 2010 Page 22 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE 0

2010 MNS SRO NRC Examination QUESTION 8 2508 eneraI Discussion KC -122 is located in the surge tank vent line and vents the tank to atmosphere. It is controlled from a local station at the surge tank by a two position, OPEN/CLOSE, pushbutton. It is normally open and receives a close signal on EMF-46A & B Trip 2 alarm. The OPEN position latches in so when the EMF signal clears, the valve will re-open. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because there are some EMFs which have control actions for a Trip 1 alarm but EMF-46 is not one of them. For example, a high radiation alarm (Trip 1) on EMF-36 (HH) will shut of the 1EMF-35/36/37 sample pump. Part 2 is plausible because 1KC-122 is unusual in that it will reopen when the high rad alarm clears. Every other valve which receives a close signal on a high rad must be manually reopened. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because there are some EMFs which have control actions for a Trip 1 alarm but EMF-46 is not one of them. For example, a high radiation alarm (Trip 1) on EMF-36 (HH) will shut of the 1EMF-35/36/37 sample pump. Part 2 is correct. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is correct. Part 2 is plausible because 1KC-122 is unusual in that it will reopen when the high rad alarm clears. Every other valve which receives a close ignal on a high rad must be manually reopened. nswer D Discussion CORRECT: See explanation above Basis for meeting the KA K/A asks for knowledge of the design features and interlocks associated with operation of the KC surge tank including associated valves. 1KC-122 is the normally open vent valve on the surge tank and the question is soliciting knowledge concerning both the design and interlocks associated with this valve. Basis for Hi Cog -_________ Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK MNS Bank Q PSSKCNO3 Development References Student References Provided Lesson Plan OP-MC-PSS-KC Pg. 21 (Rev 26) Lesson Plan OP-MC-WE-EMF page 33 (Rev 30) OP-MC-PSS-KC Obj. 10 SYSOO8 K4.02 Component Cooling Water System (CCWS)

Knowledge of CCWS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) )peration of the surge tank, including the associated valves and controls O1-9 Comments: RemarkslStatus J 401-9 Comments: No comment. Tuesday, July 13, 2010 Page 23 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination OUESTION Tuesday, July 13, 2010 Page 24 of 294

Question 8

References:

OBJECTIVES 10 Concerning the Component Cooling Water System:

     . Describe the local controls and list the indications,        x x x x including operation of the local control for KC-122.
     . Describe the control room controls and list the indications.

11 State the normal and backup sources of makeup water to the X X X X system. 12 Describe the discharge paths of the Component Cooling X X X X Water Drain Tank Pump. 13 Given a limit and/or precaution associated with an Operating X X X X X Procedure, discuss its basis and applicability. 14 Concerning AP/1/A/5500/21, Loss of Component Cooling X X X Water:

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

15 Concerning the Technical Specifications related to the Component Cooling Water System:

     . Given the LCD title, state the LCD (including any COLR            X X X values) and applicability.
     . For any LCDs that have action required within one hour,          x x x state the action.
     . Given a set of parameter values or system conditions, determine if any Tech Spec LCDs is (are) not met and any action(s) required within one hour.
    . Given a set of plant parameters or system conditions and the appropriate Tech Specs, determine required actions.             X X
    . Discuss the bases for a given Tech Spec LCD or Safety Limit.                                                              X

From Lesson Plan OP-MC-PSS-KC Pg. 21 (Rev 26) 2.6.6 ND Heat Exchanger Cooling Water Isolation Valves (KC-56 & 81).

These valves are located on the inlet of the ND Heat Exchanger and are controlled from Control Room MC-1 1. The operator must hold the open pushbutton until the valve fully opens because there is no seal-in associated with the open circuit. They are normally closed and open on a S signal. 2.6.7 ND Heat Exchanger Cooling Water Control Valves (KC-57 & 82). These valves are located in the discharge lines of the ND Heat Exchangers. It is normally controlled by flow instrumentation to maintain KC flow through the heat exchanger at 5000 gpm. They fail open on a S signal. To regain automatic control, the S and the Modulating Valves Reset must be reset. The purpose of the Modulating Valves Reset is to ensure two actions are taken prior to removing a component from its safety alignment. These valves fail in open position. Objective #10 2.6.8 KC Surge Tank Vent Valve (KC-1 22) Located in the surge tank vent line and vents the tank to atmosphere. It is controlled from a local station at the surge tank by a two position, OPENICLOSE, pushbutton. It is normally open and receives a close signal on EMF-46A & B alarm. The OPEN position latches in so when the EMF signal clears, the valve will re-open. 2.6.9 KC Surge Tank Pressure Relief (KC-972) Designed to relieve maximum water flow as a result of a ruptured NCP Thermal Barrier Heat Exchanger. Relief setpoint is 15 psig and discharges to Liquid Waste Recycle System, via Floor Drain System 2.6.10 KC Surge Tank Vacuum Relief (KC-1 23). Vacuum breaker protects the tank from collapsing in the event of a KC leak when the KC Surge Tank vent is closed. 2.6.11 Letdown Heat Exchanger Cooling Water Control Valve (KC-132). These valves are physically located in the Letdown Heat Exchanger line and regulate component cooling flow to maintain Letdown temperature at 115 °F. Valve is designed to fail open. Operation of this valve can cause changes in the NV System Demineralizers temperatures. A change in demineralizer temperature can affect the boron concentration out of the demeralizer. Decrease in temperature can cause a dilution of the NC System (cooler resin holds more boron). An increase in temperature will have the opposite effect. See OE item 5.2

From Lesson Plan OP-MC-WE-EMF page 33 (Rev 30) 2.1.11 Component Cooling Water Monitor The Component Cooling Water System is monitored by the following channels:

1(2) EMF 46A Unit 1(2) Component Cooling A

1(2) EMF 46B Unit 1(2) Component Cooling B

Objective # 2 These channels monitor the component cooling water downstream of the component cooling water coolers. 1 EMF-46A monitors heat exchanger 1A while 1 EMF-46B monitors heat exchanger 1 B. 2EMF-46A monitors heat exchanger 2A while 2EMF-46B monitors heat exchanger 2B. A radiation indication would indicate a failure of any of various heat exchangers containing primary reactor coolant or the presence of NA-24 due to sodium activation from chemical compounds (Sodium Molybdate and Sodium Tetraborate) added to the system as a corrosion inhibitor. I Objective#2,3 I Should a Trip 2 high radiation alarm be received on either IEMF-46A or IEMF 46B, the component cooling water surge tank vent 1KC122 is automatically closed to prevent release of volatile fission products. A high radiation alarm on 2EMF-46A or 2EMF-46B will automatically close 2KC122. The purpose of Auto actions: KC122 shutting will not prevent a water release to the Aux. Building. should a primary to KC leak occur, but if the leak is small it will terminate an airborne release to the Aux. Bldg. which originates in the KC System. These channels use a single range gamma liquid (Nal Scint.) detector. 2.1.12 Boron Recycle Evaporator Distillate Monitor Objective # 2 OEMF-47 Boron Recycle is used to monitor the Boron Recycle evaporator

distillate downstream of the filter. Objective # 2, 3 Normally, the distillate will be routed to the Reactor Makeup Water Storage Tanks. On a Trip 2 high radiation alarm, valve 1NB219 will divert this flow to the Boron Recycle Holdup Tank. The purpose of Auto actions are to prevents contaminating RMWST should the NB evaporator fail to function as designed. This channel uses a single range gamma liquid (Nal Scint.) detector.

Parent Question: MNS Bank PSSKCNO3 Question 62 PSSKCNO3 1 Pt Which one of the following describes the automatic operation of I KC-1 22 (KC Surge Tank Vent Valve)? A. 1 EMF-46A (B) in Trip 1 alarm will cause the vent to close; when the alarm clears the valve will automatically re-open (the OPEN position seals in). B. 1 EMF-46A (B) in Trip 2 alarm will cause the vent to close; when the alarm clears the valve will automatically re-open (the OPEN position seals in). C. 1 EMF-46A (B) in Trip 1 alarm will cause the vent to close and the CLOSE positions seals in; the valve must be locally re-opened. D. 1 EMF-46A (B) in Trip 2 alarm will cause the vent to close and the CLOSE positions seals in; the valve must be locally re-opened. Answer 62 Answer: B

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 9 SYSO 10 A 1.07 - Pressurizer Pressure Control System (PZR PCS) bility to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR PCS controls ncluding: (CFR: 41.5 / 45.5) RCS pressure Given the following conditions on Unit 1:

Repairs have just been completed on the SLIM module for I NC-29 (Pressurizer Spray Valve)
The OATC has just completed throttling I NC-29 with its controller in MANUAL to verify proper operation of the SLIM
INC-29 controller is now in AUTO After completion of testing the Pressurizer Pressure Master Controller soft controls indicate as follows:

ACTIVE F9ESSURE CONTROL PZR PRESSURE ERRR 01 PSIG MIN DEMAND MAX DEMAND SPRAY

1. What is the current demand for the Bank C heaters?

2. Atwhat PRESSURE ERROR will the Backup heaters energize?

A. 1. 17%

2. () 17 PSIG B. 1. 17%

2. () 25 PSIG C. 1. 83%

2. () 17 PSIG D. 1. 83%

2. ()25PSIG Tuesday, July 13, 2010 Page 25 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 9 2509 General Discussion C Heater Group control is always in Automatic. The SCR power controller is controlled by the Pressure Master Controller. C Heater power is rmped linearly from 0% to 100% as the Pressure Master Controller output goes from -15 psig (Error) to +l5psig (Error), regardless of system pressure. At a PZR Pressure Error of -10.0 PSIG, the Bank C Demand can be calculated as follows: 50% Demand for the Heaters occurs at a Pressure Error of 0 PSIG. 100% Demand occurs at a Pressure Error of-15 PSIG. Therefore, there is a 50% increase in Demand over a -15 PSIG change in Pressure Error. Therefore: 50/-is = x!-i0 -15x = -500 x = 33.3 (change in demand from 50%) Starting from an initial Demand of 50% the final Demand = 50% + 33.3% = 83.3% The Backup heaters energize when if the pressure error signal is -25 PSIG and de-energize when the pressure error signal increases to -17 PSIG. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part I is plausible if the applicant calculates the Demand based on a +10 Pressure Error signal as 17% Demand would be correct. Part 2 is plausible since this is the error signal at which the Backup heaters de-energize. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible if the applicant calculates the Demand based on a +10 Pressure Error signal as 17% Demand would be correct. art 2 is correct.

 .nswer C Discussion INCORRECT: See explanation above.

PLAUSIBLE: Part I is correct. Part 2 is plausible because this is the error signal at which the Backup heaters dc-energize. Answer D Discussion CORRECT: See explanation above. Basis for meeting the KA The KA is matched because the PZR PCS controls have been operated resulting in a change in NC system pressure. Based on the change in NC system pressure and the indications affected (ability to monitor) by that change (Pressurizer Pressure Error) the applicant must determine the status of the Pressurizer Heaters. Once the Operator determines what the demand for the Bank C Heaters should be, they can compare that to the actual demand signal on the soft panel to determine if the PCS is operating as required. Part of the MNS Design Basis Safety Analysis assumes that Pressurizer Pressure and Level are within normal operating limits at the beginning of a transient. For example, with regards to pressurizer pressure, part of the design basis of the Pressurizer (related to PZR pressure control) assumes that a Safety Injection does not occur on a Reactor Trip. If Pressurizer pressure is low out of the normal operating band when a Reactor Trip occurs it is possible that a Safety Injection could occur on low pressure and thus the Pressurizer would not have performed its design function with regards to maintain NC system pressure within design limits. Consequently, the ability to monitor proper operation of heaters and sprays to maintain pressure within normal limits is essential in assuring that the design limits of the system are not exceeded. Therefore, the to prevent exceeding design limits portion of the KA is also met by this question. Basis for Hi Cog This is an analysis level question because Part 1 of the question requires the applicant to calculate the C Bank Heater Demand based on the Pressurizer Pressure Error signal. Part2of the question is memory. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Tuesday, July 13, 2010 Page 26 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 9 2509

                                         -                                                 Student References Provided i.earning Objective:

1) PS-IPE-DCS #5

References:

1) Lesson Plan OP-MC-PS-IPE-DCS Section 2.4 -__________________

SYSO1O A1.07 Pressurizer Pressure Control System (PZR PCS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR PCS controls including: (CFR: 4L5 /45.5) RCS pressure 401-9 Comments: narksIStatus -____________ 401-9 Comments: No comment. Resolution / Comments: N/A Tuesday, July 13, 2010 Page 27 of 294

Question 9

References:

From Lesson Plan OP-MC-PS-IPE-DCS Section 2.4: 2.4 C Heater Group C Heater Group is made up of 7 heater banks. The heater banks have variable power control. The capacity of the C Heaters totals 484 KW. There are two power sources available for the C Heaters, LXF (normal) and LXC (Alt.). The breakers are Kirk Key interlocked so that only one can supply at a time. The supply breaker auto trips on Low PZR Level <17% and also if charging flow lowers to <10 gpm for >20 seconds, to prevent heater damage if uncovered (due to the poor heat conduction into non-liquid surroundings). When level recovers to >17% or 15 seconds after the heaters are de energized due to low charging flow and PZR level is >17%, the supply breaker must be manually reclosed, using the MCB OPEN/CLOSED control switch for the supply breaker on MC1O. With the supply breaker closed, the heaters might still not be energized, unless the SCR power controller is turned on by the pressure control system. rojective #5 C Heater Group control is always in Automatic. The SCR power controller is controlled by the Pressure Master Controller. C Heater power is ramped linearly from 0% to 100% as the Pressure Master Controller output goes from -15 psig (Error) to

+1 5psig (Error), regardless of system pressure. In the rare instance of starting from 0 psig (Error) output (2235 psig) steady state conditions (with no integral function built in),

and having a rapid step decrease in pressure with no time for integral to build in, then when the controller got to -15 psig (Error), this would be equivalent to a system pressure of 2220 psig. The same would be true for a transient in the other direction for

+15 psig (Error) and 2250 psig. There is only either 484 KW or 0 KW going to the C Heaters, with power being controlled by the percentage of time 484 KW is going to the heaters. With C Heaters 10% on, 10% of the time 484 KW is going to the heaters, and 90% of the time 0 KW is going to the heaters. There is a red indicating light on the MCB that lights during the time 484 KW is being sent to the heater. On the NC-Pressurizer and PRT DCS graphic, the C Group Heater amps will vary as the heater are energized and deenergized.

2.5 Backup Heaters Refer to Drawing 7.8, Backup Heater Control. There are three Groups of backup heaters (A, B, & D). Groups A & B have 6 Banks each. Total worth for each Group is 416 KW. Group D has 7 Banks. Total worth is 484 KW. Groups A & B have safety related power supplies (EL)(A & ELXB) and are required by Tech Specs. Group D has non-safety power supply (6 Banks from LXG, 1 Bank from SMXG at the SSF). All three Groups supply breakers trip if PZR level <17% and also if charging flow lowers to <10 gpm for >20 seconds. When level recovers to >17% the supply breakers can be

manually reclosed. Likewise 15 seconds after the heaters are de-energized due to low charging flow and PZR level is >17% the supply breakers can be manually reclosed. Unlike the C Group with its supply breaker control on MCB 1 MCi 0, the backup heaters have a different arrangement. The backup heater supply breaker controls are

on the back vertical MCB MC-5. On the front MCB MC-1 0, they have ON/OFF control that controls the M contacts in the supply breaker to actually energize/de-energize the heater Groups. Groups A & B trip on a Blackout or S. On a Blackout, they can be manually reclosed. On a S, the sequencer must be reset before they can be manually reclosed. The combined total capacity of all heaters is 1800 KW. 2.5.1 MCB Backup Heater Control Objective #5 Each Backup Heater Group has an AUTO/MAN selector switch on the MCB. In AUTO, the heaters will energize on a PZR High Level Deviation (5% > programmed level) or a PZR Low Press Deviation. The setpoint for Low Pressure Deviation are Heaters ON at -25 psig (Error) output on the Pressure Master Controller and Heaters OFF at -17 psig (Error) output. The reason for energizing heaters on a high level deviation is to warm the liquid temperature back to saturation on the assumed cold water insurge that caused the high level. In MAN, the heaters will energize via the MCB ON/OFF control switch, one for each backup heater. When in manual, the AUTO functions are disabled (still get the PZR Low Low Level, Blackout, and charging flow <10 gpm for> 20 seconds, & S trips, as appropriate). There is indication of Backup Heater status and amperage supplied to each Heater Group on NC Pressurizer and PRT DCS Graphic.

2.5.2 Local Heater Control A & B Group Heaters can be locally controlled from the ASP. This is accomplished by going to LOCAL on the CR/LOCAL switches. In LOCAL, the MCB functions are disabled (Auto energize & MAN operation). A & B Group heaters would still trip on a Blackout or S, but not on PZR Low-Low level (17%). Bank 1 of D Group can be locally controlled at the SSF. This is accomplished by going to LOCAL on the REM/LOCAL switch. In LOCAL, the MCB functions are disabled (Auto energize & MAN operation). Bank 1 is energized via the ON/OFF switch at the SSF. 2.6 Pressurizer Spray Control Flow to the spray nozzles (900 gpm maximum capacity) is controlled by the positioning of valves NC-27 & NC-29. Each spray valve has a SLIMs controller on the MCB. Refer to Drawing 7.4. The controller sends a 0 100% output through an I/P converter for the

resulting 3 -15 psig pneumatic signal to control the air operated spray valve (3 psig full

closed, 15 psig full open). When the MCB Spray controller is in MAN, the Operator

can use the raise/lower pushbuttons to position the output to the desired value. When the MCB Spray Controller is in AUTO, the Pressure Master Controller controls the Spray Controller output. Objective #5 The Spray Controller output is ramped linearly from 0% 100% as the Pressure Master

Controller output goes from +25 psig (Error) to +75 psig (Error). Positive feedback of spray valve position (OPEN, INTERMEDIATE, or CLOSED) is provided via illuminated windows on the PV bar graph on the spray controller (Soft Control and SLIMs). These lights are generated from signals received from the valve limit switches. When the full CLOSED limit switch is made up, the bottom window will be the only window that is lit.

When the valve comes off the full CLOSED limit switch the middle window will illuminate and now both the bottom and middle windows will be lit. When the Valve reaches the full open position and the full OPEN limit switch is made up the top window will illuminate. At this point all windows, bottom, middle, and top will all be lit.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 10 25lO SYSOI2 A2.02 - Reactor Protection System (RPS) ility to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to

,rrect, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 /43.5 /45.3 /45.5)

Loss of instrument power Given the following conditions and sequence of events on Unit 1:

The unit is operating at 100% RTP
The crew enters AP-016 (Malfunction of Nuclear Instrumentation) due to N-42 lower detector failing LOW
IAE has placed the required bistables in the trip condition per AP-016
A complete loss of 1EKVA occurs Which ONE (1) of the following lists the required procedure flowpath forthese conditions?

A. Continue in AP-016 B. Enter AP-003 (Load Rejection) C. Enter E-0 (Reactor Trip or Safety Injection) D. Enter AP-015 (Loss of Vital or Aux Control Power) Tuesday, July 13, 2010 Page 28 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 10 C General Discussion -____________________ With N-42 failed, OTDT picks up. Also, since required bistables are tripped, all Reactor Protection bistables for that channel (Channel II) are tripped. Loss of a vital bus will cause that channels bistables (Channel I) to pick up. This results in a 2/4 logic on the protection bistables resulting_in a runback and reactor trip. Answer A Discussion INCORRECT: See explanation above. USIBLE: This answer is plausible since AP-016 will address the N42 failure but does not take priority. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible since there are 2/4 OTDT and if a reactor trip did not occur, a runback would. --_______ Answer C Discussion CORRECT: See explanation above. Answer DDiscussion - iECT: See explanation above. PLAUSIBLE: This answer is plausible since AP-15 will address the IEKVA failure but does not take priority. Basis for meeting the KA The KA is matched because the applicant must determine the effect of the loss of instrument power on the Reactor Protection System and determine the appropriate procedure based on that malfunction. The predict part of the KA is met in that the applicant must determine from analyzing the given conditions that the reactor has tripped which effects the procedure flowpath that must be taken. Basis for Hi Cog ( Riigher cognitive level question because the applicant must associate multiple pieces of information to determine the correct answer. rst, the applicant must recall that on a loss of IEKVA, power is lost to Nuclear Instrument Channel I. The applicant is given that IAE has ced the required bistables for Power Range Channel II (N-42) in the tripped condition. The applicant must recall from memory that placing

 ,ie required bistables in the tripped condition means all protection bistables for that channel and must also recall that loss of power to Nuclear Instrument Channel I will de-energize Control Power to N-4lresulting in a trip of the protection bistables for that channel. The applicant must ecafltheRPS protection logic for Nis to determine that a Runback and Reactor Trip should have occurred.

Basis for SRO only -____________________________________ Eb_Level Cognitive Level QuestionType Question Source RO Comprehension BANK 2008 NRC Q37 (Bank 543) rólopment References Student References Provided ion Plan OP-MC-IC-ENB Section 2.7 SYSO12 A2.02 Reactor Protection System (RPS)

Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; 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 /45.5) Loss of instrument power 401-9 Comments: -- ;RemarkslStatus - - 40 1-9 Comments: Change what to Which in the WOOTF statement. Please verii that B is in fact not a correct answer in and of itself. I suggest that B be replaced just to remove all doubt based on B distractor analysis. This Q is a U based on possible 2 correct answers. FAC re verify. If B is found to be acceptable, U will become an S. Tuesday, July 13, 2010 Page 29 of 294

___ FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 10 Resolution / Comments: Answer B cannot be a correct answer. A reactor trip WILL occur for the conditions given and the correct procedure flowpath is to transition to E-O. Changed what to which in the stem per Lead Examiners recommendation. See attached file for revised copy of question. Tuesday, July 13, 2010 Page 30 of 294

Question 10

References:

From Lesson Plan OP-MC-IC-ENB Section 2.7: 2.7 Power Supplies NIS Channel I EKVA NIS Channel II EKVB NIS Channel III EKVC NIS Channel IV EKVD Wide Range Train A EKVA Wide Range Train B EKVD 3.0 SYSTEM OPERATION 3.1 Normal Operation 3.1.1 Operating Procedures The Excore Nuclear Instrumentation System provides the operator with neutron flux indication for all modes of operations. During each reactor startup, a healthy skepticism concerning the validity of power indications is warranted, particularly following a refueling outage. Changes in plant equipment or conditions, along with a strong desire to return the plant to full operation, may influence personnel to accept less than complete explanations for discrepant indications. For example, excessive electrical generation for the nuclear power indicated (a symptom of miscalibrated nuclear instruments) has been attributed to factors such as: cold circulating water temperature, expected efficiency improvements, and changes in core design or instrumentation.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2008 CNS RO NRC Examination QUESTION 37 543 KA_desc SYSO12 Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; 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 A2XJ2 / 45.5)ElLoss of instrument power Given the following conditions and sequence of events:

Unit 1 was operating at 100% power.
The crew has entered AP/1/A/5500/016 (Malfunction of Nuclear Instrumentation System) due to N-42 lower detector failing LOW
IAE has not yet placed the required bistables in the trip condition per AP/1 1A155001016.
A complete loss of 1ERPD occurs What procedure takes priority for these conditions?

A. Continue in AP/1/A15500/01 6 B. Enter APIIIA/55001029 (Loss of Vital or Aux Control Power) C. Enter AP/1/A/5500/003 (Load Rejection) D. Enter EP/1/A/5000/E-0 (Reactor Trip or Safety Injection) Monday, February 08, 2010 Page 73 of 150

FOR REVIEW ONLY DO NOT DISTRIBUTE 0

2008 CNS RO NRC Examination QUESTION 37 General Discussion With N-42 failed, OTDT picks up. Loss of a vital bus will cause that channels bistables to pick up (in general) including OTDT. This causes a 2/4 situation on the OTDT runback and reactor trip. Ran on simulator at BOL and EOL and confirmed that lower detector only failing low would cause OTDT bistable to switch state. Answer A Discussion [This procedure will address the N42 failure but does not take priority. Answer B Discussion This procdure will address ERPD failure but does not take priority. Answer C Discussion There are 2/4 OTDT and if a reactor trip did not occur, a runback would. Answer D Discussion CORRECT. Basis for meeting the KA

Basis for Hi Cog
Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK Development References Student References Provided ENB lesson EPL lesson KA KA_desc SYSO 12 Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; 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 A2.02 / 45.5)E:Loss of instrument power 401-9 Comments: RemarkslStatus D12A2.02 Question appears to match K/A. NEW Will the OTDT bistables be in ifjust the lower detector on N-42 failed

[ow? If not, and the bistables have not been placed in a tripped condition, the reactor may not trip, and C would be correct. Please explain. JEW Monday, February 08, 2010 Page 74 of 150

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 11 E 25ll SYSO13 K2.O1 Engineered Safety Features Actuation System (ESFAS)

nowledge of bus power supplies to the following: (CFR: 41.7) £SFAS/safeguards equipment control Given the following conditions on Unit 1:

A Small-Break LOCA has occurred
The crew has reached the step in E-1 (Loss of Reactor or Secondary Coolant) to reset SI and the Sequencers
The crew is unable to reset the Sequencers Which ONE (1) of the following describes the locations where Operators must be dispatched to de-energize BOTH Sequencers?

A. IEVDA;1EVDB B. 1EVDA;1EVDD C. 1EVDB;1EVDC D. 1EVDC;1EVDD Tuesday, July 13, 2010 Page 31 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 11 eneral Discussion -- If one or both Sequencers can not be reset, E-l Step 8 RNO directs operators to be dispatched to the 125 VDC Vital Instrument and Control Panelboard to de-engerize the affected Sequencer. The Train A sequencer is power from 1EVDA Breaker 6 and the Train B Sequencer is power from 1EVDD Breaker 8. Answer A Discussion INCORRECT: See explanation above. LUSlBLE: This answer is plausible because 1EVDA is correct and IEVDB is another I25VDC Vital Instrument and Control PaneIboaJ Answer B Discussion iRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because 1EVDB and 1EVDC are both 125 VDC Vital Instrument and Control Panelboards. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because IEVDD is correct and 1EVDC is another I25VDC Vital Instrument and Control Panelboard. Basis for meeting the KA The KA is matched because the DO Load Sequencers control ESFAS/Safeguards equiment. The applicant must know the power supplies i1 DO Load Sequencers to know where to dispatch_NEOs to_de-energize_the_sequencers. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW - ontferences Student References Provided Lesson Plan OP-MC-DG-EQB Section 2.4 L/h/A/5500/E-1 (Loss or Reactor or Secondary Coolant) SYSO13 K2.01 Engineered Safety Features Actuation System (ESFAS)

Knowledge of bus power supplies to the following: (CFR: 41.7) ESFAS/safeguards equipment control iCoents: RemarkslStatus 401-9 Comments: No comment. Resolution I Comments: N/A Tuesday, July 13, 2010 Page 32 of 294

Question 11

References:

From Lesson Plan OP-MC-DG-EQB: Event Recorder Inputs Sequencer A actuated LOCA (ER1 79) Sequencer B actuated LOCA (ER202) 1 ETA loss of voltage Phase X, Y, or Z (3 1 ETB loss of voltage Phase X, Y, or Z (3 points) (ER187, 188, 189) points) (ER210, 211, 212) Train A Load group 1-10 energized (11 Train B Load group 1-10 energized (11 points) (ER191, 192, 193, 194, 195, 196, points) (ER214, 215, 216, 217, 218, 219, 197, 198, 199, 200, 201) 220, 221, 222, 223, 224) Train A Accelerated sequence on (ER185) 1ETB load shed (ER213) 1 ETA load shed (ER1 90) Auto reset sequencer B (ER204) Auto reset sequencer A (ER181) Train B Blackout logic initiated (ER205) Train A Blackout logic initiated (ER182) Train B Blackout logic actuated (ER206) Train A Blackout logic actuated (ER1 83) Train B Accelerated sequence on (ER208) Train A 8 sec. UV test complete (ER1 84) Train B 8 sec. UV test complete (ER207) Sequencer A reset actuated (ER1 80) Sequencer B reset actuated (ER203) D/G A start enabled (ER186) D/G B start enabled (ER209) D/G A Committed Time Sequence DIG B Committed Time Sequence Commenced (ER425) Commenced (ER426) ETA Degraded Voltage Phase X, Y, or Z ETB Degraded Voltage Phase X, Y, or Z (3 Points) (ER445, 446, 447) (3 Points) (ER450, 451, 452) ETA Degraded Voltage Alarm Timer (ER ETB Degraded Voltage Alarm Timer (ER 448) 453) ETA Degraded Voltage Trip (ER449) ETB Degraded Voltage Trip (ER454) Power Supplies The Diesel Generator Load Sequencer System is powered from the 125 VDC Vital Instrumentation and Control System. (Train IA 1 EVDA, Train 1 B 1 EVDD).

                                             -                   -

Each sequencer cabinet has a space heater. These space heaters are powered from local non-safety lighting panelboards.

From E-1 (Loss of Reactor or Secondary Coolant): RESONSE PEPONSE NOT O3TAINEO AT:oN/E:?E::TEt

8. (Continued)

g. Reset the following:

1) S/I. 1) Reset S/I PER EP/1/A/5000/G-1 (Generic Enclosures), Enclosure 23 (Local Reset of S/I Signal).

2) Sequencers. 2) Dispatch operator to open affected sequencer control power breaker:

A Train - 1 EVDA Breaker 6
B Train - IEVDD Breaker 8.

3) Containment spray.

h. IF AT ANY TIME a B/C signal occurs, THEN restart S/I equipment previously on.

i. Stop NS pumps.

j. Close the following:

1NS-29A (1A NS Hx Outlet Cont Outside Isol)
1NS-32A(lANSHxOutletCont Outside Isol)
1NS-15B (lB NS Hx Outlet Cont Outside Isol)
1NS-12B (lB NS Hx Outlet Cont Outside Isol).

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 12 2.4.45 Containment Cooling System (CCS)

SYS022 YS022 GENERIC Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10/43.5/45.3/45.12) Given the following conditions on Unit 1:

The unit is operating at 100% RTP
A small NC System leak occurs inside Containment
Annunciator 1AD-9 /A8, (CONT .5 PSIG ALERT) is received Which ONE (1) of the following is an expected response of the VL AHUs and the Containment Pipe Tunnel Booster Fans (PTBFs)?

A. All VL AHU(s) start and shift to HIGH speed Both PTBFs start and shift to HIGH speed B. All VL AHU(s) start and shift to HIGH speed The PTBFs are running according to their switch positions C. Operating VLAHU(s) shift to HIGH speed, Idle fans remain OFF Both PTBFs start and shift to HIGH speed D. Operating VL AHU(s) shift to HIGH speed, Idle fans remain OFF The PTBFs are running according to their switch positions Tuesday, July 13, 2010 Page 33 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 12 2512 neraI Discussion At the initiation of the 0.5 psig lower Containment pressure signal, all four VL air handling units and both Pipe Tunnel Booster Fans will start and switch to HI speed. HVAC switch control is regained when pressure is less than 0.5 psig. Answer A Discussion CORRECT: See explanation above. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: The first part of this distracter is correct. The second part concerning the PTBFs is plausible because this would be the correct response for an SI signal. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: The first part of this distracter is plausible because the VL AHUs do shift to high speed, and if the applicant confuses the response of other VUL components (VR, S!G Booster Fans and VR fans) which will only respond if they are selected to be running. The second part concerning the PTBFs is correct. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: The first part of this distracter is plausible because the VL AHUs do shift to high speed, and if the applicant confuses the response of other VUL components (VR, S/G Booster Fans and VR fans) which will only respond if they are selected to be running. The second part concerning the PTBFs is plausible or reasons stated above. _,_Basis for meeting the KA - ( is K/A is a generic applied to the containment cooling system. The question requires the applicant to possess the ability to interpret the gnificance of an annunciator associated with the CCS by indentifiing what effect the alarm will have on the system. There are very few

annunciators associated with the CCS system and the only time they would present an opportunity to Prioritize would be if containment pressure was approaching 1 psig setpoint which would result in an SI. Dealing with the SI would then become the priority and the system part of the K/A would not be met.

Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Lesson Plan OP-MC-CNT-VUL Page 31 (Rev 28) OP-MC-CNT-VUL Obj. 4 SYSO22 2.4.45 Containment Cooling System (CCS)

SY5022 GENERIC Ability to prioritize and interpret the significance of each annunciator or alarm. (CFR: 41.10 / 43.5 / 45.3 / 45.12) RemarkslStatus s: t cCoen 9 Comments: No comment. Resolution / Comments: Tuesday, July 13, 2010 Page 34 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 12 2512 N/A Tuesday, July 13, 2010 Page 35 of 294

Question 12

References:

OP-MC-CNT-VUL Obj. 4 OBJECTIVES N N L L L L L P P0 OBJECTIVE 0 0 R S R R 00

1. State the purpose of the following Containment Ventilation X X X X Subsystems

   . Upper Containment Ventilation System.
   . Lower Containment Ventilation System.
   . Control Rod Drive Ventilation System.
   . Incore Instrumentation Room Ventilation System.

2. State the source of cooling water to the upper and lower X X X X containment ventilation units.

3. Discuss the operation of the Containment Ventilation Systems X X X X (VU,VL,VR,VT) including the components operating during normal unit operations.

4. State the automatic actions that occur to the Lower X X X X X Containment Ventilation units if containment pressure increases to 0.5 psig.

5. Discuss the automatic alignment of the Containment X X X X X Ventilation Systems (VU, VL, VR, VT) following a:

   . Safety Injection signal.
   . Blackout signal.

6. Concerning the Reset/Retransfer switches: X X X X X

   . List the units having a Reset/Retransfer switch.
   . Discuss the purpose and operation of the switch.

7. Describe the local controls and indications associated with the X X X X X Containment Ventilation Systems.

8. Describe the Control Room controls and indications X X X associated with the Containment Ventilation Systems.

From Lesson Plan OP-MC-CNT-VUL Page 31 (Rev 28) 1.0 SYSTEM OPERATION 1.1. Normal Operation VL System Operation Objective #3 I Typical configurations for operation of the VL ventilation units are listed as follows in order of increasing cooling capacity:

1. Two to four units at low speed

2. Three units at high speed with one standby unit, and

3. Four units at high speed.

The number of ventilation units needed to cool lower Containment depends upon the season of the year, the cooling water inlet temperature, and the Containment heat load. The lower containment heat load has decreased due to improvements in insulation techniques. Therefore, operation with only two VL AHUs in low speed is now possible. The most desirable configuration for operation of the ventilation units is low speed operation. This will minimize the wear and required maintenance on the units. Optimum VL AHU and RV Pump configuration is based on Lower Containment Weighted Average Temperature (LCWAT), the number of VL AHUs in operation and the speed the VL AHUs are operating in. In Modes 1 through 5, RN is the preferred source of cooling water. RV pumps can supply cooling, but are not the preferred source. In Mode 6, or No Mode, cooling water is not required. Objective #4 At the initiation of the 0.5 psig lower Containment pressure signal, all four VL air handling units and both Pipe Tunnel Booster Fans will start and switch to HI speed. HVAC switch control is regained when pressure is less than 0.5 psig. The pressurizer booster fans have an electric interlock so that both fans can not be operated at the same time. One fan is operated during normal operation. One pipe tunnel booster fan is operated in High speed during normal operation. Each steam generator area booster fan operates during normal operation. VR System Operation Objective #3 Normal operation will consist of running a minimum of three (3) of the four (4) VR fans.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 13 2513 SYSO22 K1.O1 Containment Cooling System (CCS)

iowledge of the physical connections and/or cause-effect relationships between the CCS and the following systems: (CFR: 41.2 to 41.9/45.7 45.8) SWS/cooling system Which ONE (1) of the following describes the operation of the RV System if Containment pressure reaches 2.8 PSIG? A. The RV Containment isolation valves will Auto Close on the (ST) signal. Containment cooling will be provided to the RN non-essential header. B. The RV Containment isolation valves will Auto Close on the (Ss) signal. Containment cooling will be provided to the RN non-essential header. C. The RV header is isolated from the RN header by the (ST) signal. The RV pumps will Auto Start on RN non-essential header low pressure to supply the Containment AHUs. D. The RV header is isolated from the RN header by the (Ss) signal. The RV pumps will Auto Start on RN non-essential header low pressure to supply the Containment AHUs. Tuesday, July 13, 2010 Page 36 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 13 25l3 General Discussion In this question, a 1 psig pressure in containment would result in a Safety Injection. As a result of the SI a (SS) Safety Injection signal and a (ST) Phase A Isolation signal would be generated. The normal supply to Containment cooling is the RN non-essential header. This header is supplemented by flow from the RV (Containment Cooling) pumps via a tie into the RN piping from RV downstream of RN-42A. RN-42 closes on a (SS) signal resulting is isolation of RN to the Containment Ventilation portion of the RN Essential header which isolated RN from RV. This would result in lower pressure of this header and when the pressure reaches 50 psig any RV pump selected to Auto will start. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: The first part is plausible if the applicant confuses the signal that closes the RV containment isolation valves. The RV containment isolation valves to get an ESF signal to close. However, they close on a Phase B (SP) signal instead of a Phase A signal. Since RN is the normal supply for Containment cooling the second part is plausible as well. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: The first part is plausible if the applicant confuses the signal that closes the RV containment isolation valves. The RV containment isolation valves to get an ESF signal to close and many containment isolation valves do close on a Safety Injection signal. However, they close on a Phase B (SP) signal instead of a Safety Injection signal. Since RN is the normal supply for Containment cooling the second part is plausible as well. Answer C Discussion INCORRECT: See explanation above. LAUSIBLE: The first part of this answer is plausible if the applicant confuses what signal isolates the RV header from the RN header. The RV neader is isolated from the RN header on a Safety Injection and since the SI signal also generates a Phase A signal, it is plausible for the applicant to conclude that the Phase A signal cause the isolation instead of the Safety Injection signal. The second part is correct. Answer D Discussion CORRECT: See explanation above. Basis for meeting the KA This K/A is met because the applicant is required to recall knowledge that involves an understanding of the physical connections between RV and RN. In the question a phase A has occurred which only effects RN but due to how the two systems are tied together, RV is ultimately effected and to correctly answer the question an understanding of the relationship is required. In this way, this question also meets the cause-effect angle of this K/A. RV is our containment cooling system. Basis for I-li Cog rThis is a hi cog question because it involves a level of analysis of given situation, apply system knowledge and solve a problem of what both would be the effect and how the system would respond to the conditions given in the stem. ______________ Basis for SRO only - Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK MNS Bank Q CNTRVO28 Development References Student References Provided Lesson Plan OP-MC-CNT-RV Pg. 27 JP-MC-CNT-RV Objs 2,4 & 13 SYSO22 Kl.Ol - Containment Cooling System (CCS) Knowledge of the physical connections and/or cause-effect relationships between the CCS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8) Tuesday, July 13, 2010 Page 37 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 13 2513 SWS/cooling system

.01-9 Comments:       -             Remarks!Status                                                  -

401-9 Comments: 1 believe that it is common knowledge that Containment isolation occurs on a phase B isolation. I believe that distractors A and B are NP. Consider replacing or modif,ing or putting something in the stem to give them false credibility. This Q is U due to potentially 2 NP distractors, Resolution / Comments: In this particular case, it is true that the RV Containment isolation valves close on a Phase B isolation. However, the majority of all Containment Isolation valves close on a Phase A signal. Therefore, it is plausible for the applicant to conclude that the RV Containment isolation valves would close on a Phase A signal (ST) or on a Safety Injection (SS) signal which causes a Phase A signal. Tuesday, July 13, 2010 Page 38 of 294

Question 13

References:

OBJECTIVES s NNLLL E OBJECTIVE L L P P 0

,.                                                              OORSR R0O 1 State the purpose of the Containment Ventilation Cooling     X Water System.

2 Describe the flowpath of water from the RV pump suction to X the RN header discharge. 3 Describe the RV pump strainer design. X 4 List the signals that will AUTO-START a RV pump. X X 5 Describe how the RV pumps are protected during low flow X X conditions. 6 List the loads supplied by the RV header. X 7 List the safety related components associated with the RV X X system. 8 List the signal that will Auto-Close RN-3O1AC and RN-302B X X (RV pump suction supply header isolation valves). 9 Describe the RV pump instrumentation and controls. X 10 Describe the normal operation of the RV System, including X X Manual operation of the RV pumps. 11 Describe the sequence to swap and clean the RV pump X X suction strainer if RV pumps are operating. 12 Given a Limit and/or Precaution associated with an operating X X procedure, discuss its basis and applicability. 13 Explain the status of the RV pumps during normal and X abnormal operation.

From Lesson Plan OP-MC-CNT-RV Pg 27 (Rev 19) 3.2 Abnormal and Emergency Operation Objective # 13 The RV System is not required during emergency operations or for the safe shutdown of the plant. It may be desirable, though not required, for the RN System to support the Auxiliary and Reactor Building AHUs for some emergency operating scenarios. For a Loss of Offsite Power (LOOP) Event, the RN System can provide cooling water to the RV loads as RN receives emergency power from the emergency diesel generators. Isolation valve RN42A remains open to allow RN Train A water to supply RV loads since the RV pumps may be inoperable upon a loss of offsite power. Upon a Containment High-High Pressure Signal (SP), the RV Containment Isolation Valves will close, thus isolating the Upper Containment, Lower Containment, and lncore Instrumentation Room AHUs from the RV System. The RV pumps are also isolated from the Low Level Intake when the RN valves respond to the SP signal. The RV System has no change of alignment or operation on a Containment High Pressure Signal (SS). However, the RN System does have significant alignment and operation responses to the SS Signal that could affect RV. RN-42 closes on an SS signal, leaving RV to supply VL, VU, and VT. Other RN responses could also lower the pressure in its non-essential header, thus activating any RV pump aligned in the AUTO mode. For this reason, all three RV pumps should not be aligned in the MANUAL mode at the same time unless at least one pump is running.

Parent Question MNS Bank Question 27 CNTRVO28 I Pt Which ONE of the following describes the operation of the RV system upon receipt of a Containment High Pressure (S) signal? 5 A. The RV Containment isolation valves will close as well as the Low Level Intake suction isolation valves. B. The RV pump in auto is started on the Ss signal to supply the header. The RV header is isolated from the RN header by the SS signal. C. The RV pump in auto is started on the S3 signal. The combined flow of the RV and RN system operate together to supply a higher flow rate to the Containment AHUs. D. The RV system is isolated from the RN system on the Ss signal and the RV pump in auto will start on low pressure. Answer 27 D

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 14 ..SYSO59 A2.06 Main Feedwater (MFW) System

bility to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to orrect, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 /43.5 /45.3 /45.13) Loss of steam flow to MFW system Given the following conditions on Unit 1:

The unit is increasing power after a forced shutdown
0P111A161001003 (Controlling Procedure for Unit Operation) is in effect
1A CF pump is in service
When 40% RTP is reached, the crew closes 1HM-95 (AS to A&B CF Pumps)
When IHM-95 closes, the crew observes the IA CF pump speed and DIP decreasing and FRVs opening

1. What is the cause of the indications described above?

2. What action is required to continue the power increase?

A. I. ISP-I (SM to CF Pump 1A) is closed.

2. Dispatch an operator to open ISP-I. Main Steam is the primary supply to the CF pumps between 20% and 80% RTP.

B. I. ISP-I (SM to CF Pump 1A) is closed.

2. Dispatch an operator to open 1 SP-1. Main Steam is the primary supply to the CF pumps between 20% and 100% RTP.

C. I. MSR cross over steam pressure is inadequate.

2. IHM-95 must be reopened. MSR crossover steam is the primary supply to the CF pumps between 40% and 80% RTP.

D. I. MSR cross over steam pressure is inadequate.

2. 1HM-95 must be reopened. MSR crossover steam is the primary supply to the CF pumps between 40% and 100% RTP.

Tuesday, July 13, 2010 Page 39 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 14 2514 General Discussion - During a unit start up, SP-1(2) are opened at approximately 15% RIP. This is done to align the HP steam supply to the CF pumps. When the unit reaches 40% RIP the procedure directs the crew to close 1HM-95 which isolates the Aux Steam header from the CF pumps. At this point the LP steam supply is provided by MSR crossover steam which, at 40%, is at too low a pressure to provide much flow. The majority of steam is supplied by HP steam via SP-1(2). The low pressure governor valve opens first and is supplied by the Auxiliary Steam System until the Moisture Separators Reheater (MSR) steam has sufficient capacity to supply which occurs above 80% power. The high-pressure governor is supplied by the Main Steam System and is used when the low-pressure governor is not able to meet the demand. High-pressure steam will be supplied automatically if low-pressure stream cannot maintain turbine speed. At 100% RIP the FWPT steam supply is from MSR exhaust only with the HP governor valves completely closed. With the given power level, closure of ISP-I would have no effect on FWPI operation. Aux Steam is [ally isolated at 100% power but is the primary steam supply at low power levels Answer A Discussion [3iRECI: See explanation above. Answer B Discussion iEORRECT: See explanation above. PLAUSIBLE: Part (1) is correct and therefore plausible Part (2) is plausible because the action described is correct and HP steam is aligned to the CF pump all the way to 100%. Since Main Steam pressure is higher than MSR crossover pressure it would be reasonable for the applicant to pick this as the primary supply. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: Part (1) is plausible because procedurally. 1HM-95 is closed at 40% RTP which isolates the Aux Steam Supply. This connection ties in to the Feed pump in parallel with the MSR crossover supply. It would be reasonable for the applicant to misinterpret this alignment to mean that the AS supply is being replaced with the crossover steam supply and the replacement source should be at a high enough pressure to upply the pumps. rart (2) is plausible because reopening IHM-95 would restore proper steam pressure to the feed pumps. However, the combination of Aux Steam and MSR crossover steam will not be sufficient to allow the power increase to continue. SP-1 must be opened or as power is increased the CF pump speed and D/P will again decrease due to insufficient steam supply. The range given for the MSR supply is consistant with where the isolation of AS is taking place. Since Main Steam pressure is higher than MSR crossover pressure it would be reasonable for the applicant to pick this as the primary supply above 80% Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: Part (1) is plausible because procedurally, 1HM-95 is closed at 40% RTP which isolates the Aux Steam Supply. This connection. ties in to the Feed pump in parallel with the MSR crossover supply. It would be reasonable for the applicant to misinterpret this alignment to mean that the AS supply is being replaced with the crossover steam supply and the replacement source should be at a high enough pressure to supply the pumps. Part (2) is plausible because reopening IHM-95 would restore proper steam pressure to the feed pumps. However, the combination of Aux Steam and MSR crossover steam will not be sufficient to allow the power increase to continue. SP-l must be opened or as power is increased the CF pump speed and D/P will again decrease due to insufficient steam supply. MSR crossover pressure is the primary supply to the CF pumps between 80% and 100% it would be reasonable for the applicant to pick 40% because this is consistent with where the isolation of AS is taking place. Basis for meeting the KA The K/A is matched because the applicant is presented with a scenario where adequate steam flow to the operating CF has been lost. (Loss of steam flow to the MFW system) He is then presented with a series of plausible failures and asked to predict if a given condition would result in the condition described in the stem of the question along with what would be required to mitigate the consequences of the stated condition. (A2 K/A is addessed) Basis for Hi Cog This is a hi cog question because it involves a level of analysis of given situation, applying system and operational knowledge to predict an itcome. asis for SRO only Tuesday, July 13, 2010 Page 40 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 14 __J. Job Level Cognitive Level QuestionType Question Source ( RO Comprehension NEW Development References Student References Provided Lesson Plan OP-MC-MT-MSR pages 12 and 13: Plan OP-MC-CF-CF page 17: MC-MT-MSR Obj. 2 SYSO59 A2.06 Main Feedwater (MFW) System

Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; 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 /45.13) Loss of steam flow to MFW system 401-9 Comments: RemarkslStatus 1-9 Comments: No comment. Resolution / Comments: N/A Tuesday, July 13, 2010 Page 41 of 294

Question 14

References:

OP-MC-MT-MSR Obj. 2 2 Describe the flowpaths for the Moisture Separator Reheaters x x x x including the following:

     . MSR shell side.
     . First Stage Reheater (Low Pressure) and Drain System.
     . Second Stage Reheater (High Pressure) and Drain System.
     . FWPTAandB.

From Lesson Plan OP-MC-MT-MSR pages 12 and 13:

1.0 INTRODUCTION

1.1 Purpose Objective # 1 The Moisture Separator Reheater (MSR) System is designed to take the high pressure turbine exhaust, remove the entrained moisture, provide heating through the first and second stage reheaters and supply moisture free, superheated steam to the low pressure turbines to improve efficiency and reduce maintenance on the low pressure turbine blading by reducing the low pressure turbine exhaust moisture. The MSR shell and the first and second stage reheater tube bundle drains are then returned to the condensate cycle, which improves cycle efficiency. 1.2 General Description At full power, exhaust steam exits the high pressure turbine at about 176 psia and 14% moisture content and flows to the Moisture Separator Reheaters (MSRs). The steam is first passed through a moisture (chevron) separator where approximately 10 percent of the flow is extracted as moisture and drained to a drain tank. The remaining 90 percent flows up through a two-stage steam-heated reheater where steam quality is increased and temperature is raised to approximately 150°F superheat. From high pressure turbine exhaust to low pressure turbine inlet, there is a pressure loss of approximately 8 to 9 psi at full power. Objective # 2 A steam supply is provided for operation of the Turbine Driven Main Feedwater Pumps from the reheated steam prior to entering the low pressure turbines. Once turbine load is approximately 80%, the steam exiting Al and B1 MSRs is the source of steam for the main feedwater pump turbines through the LP stop/governor valves. 2.0 COMPONENT DESCRIPTION Objective # 6 In order to prevent turbine overspeed as a result of backflow or flashback, the first stage steam supply from A heater bleed, the MSR drain tank inlets and outlets and the first and second stage drain tank outlets are equipped with piston operated check valves. There are different types of these valves used in the MSR system. One type, when supplied with air (open demand) a piston moves to compress the spring and fully open the valve. The valve is held in the open position. If flow were to reverse, the valve would close against actuator air pressure.

From Lesson Plan OP-MC-CF-CF page 17:

1.0 INTRODUCTION

1.1 Purpose Objective # I The purpose of the Main Feedwater system is to take treated Condensate (CM) System water, heat it further to improve the plants thermal efficiency, and deliver it at the required flow rate, pressure and temperature to the steam generators. The CF System is designed to maintain proper S!G water levels with respect to reactor power output and turbine steam requirements The CF System provides feedwater isolation (FWI) to containment if a FWI signal is generated. 1.2 General Description Objective # 2 Student will be required to draw a simplified system diagram as shown on Drawing 7.1. The Feedwater System begins at the Main Feedwater (CE) Pump suction header. The CF pumps discharge to the High Pressure Heaters (Al, A2, A3 and Bl, B2, B3) where reclaimed steam from the Moisture Separator Heaters and High Pressure Turbine extraction steam is used to increase feedwater temperature from 360°F to 440°F. The flow continues from the HP heaters through the feedwater control valves, containment isolation valves to the steam generators. The steam generators are used to produce steam for use in the main turbine and other auxiliary loads. 2.0 COMPONENT DESCRIPTION 2.1 Main Feedwater Pumps Objective # There are two 50% capacity feedwater pumps driven by two 50% capacity variable speed turbines (refer to Drawing 7.2). The main feed pumps increase system pressure from approx. 400 psig at its suction to approx. 1200 psig at its discharge at 100% power. High and low-pressure governor valves control the turbine speed. The low pressure governor valve opens first and is supplied by the Auxiliary Steam System until the Moisture Separators Reheater (MSR) steam has sufficient capacity to supply which occurs above 80% power. The high-pressure governor is supplied by the Main Steam System and is used when the low-pressure governor is not able to meet the demand. High-pressure steam will be supplied automatically if low-pressure stream can not maintain turbine speed. A check valve is provided in the low-pressure supply to

prevent reverse flow from the high-pressure turbine. For more information on the Main Feedwater Pump Speed control, refer to lesson plan OP-MC-CF-IWE.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 15 2515 SYSO25 K6.O1 Ice Condenser System

owledge of the effect of a loss or malfunction of the following will have on the ice condenser system: (CFR: 41.7/45.7) jpper and lower doors of the ice condenser Given the following conditions on Unit 1:

The unit is in MODE I at 10% RTP
IAD-9 /A5 (ICE COND LOWER IN LET DOORS OPEN) alarm is LIT
The lower inlet door position display panel indicates that a door is open
The door is confirmed to be cracked opened. The door will not move further open and cannot be closed
No other alarms related to the ice condenser, NE system orAHUs are lit Which of the following is REQUIRED to be entered based on the current plant conditions?

1. Tech Spec 3.6.13 Ice Condenser Doors

2. Tech Spec 3.6.12 Ice Bed

3. Selected Licensee Commitment 16.6-3 Ice Condenser Door Position Monitoring System A. 1 ONLY B. 1 and 2 ONLY C. I and 3 ONLY D. 1,2,and3 Tuesday, July 13, 2010 Page 42 of 294

_______ ________ ________ ________ ________ ______ ______ _____ _______ _____ FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 15 2515

   .jeneral Discussion Per TS 3.6.13 (Ice Condenser Doors) The ice condenser inlet doors, intermediate deck doors, and top deck doors shall be OPERABLE and closed  while in MODES       1, 2,    and 4.

3, Answer A Discussion [2RT: See explanation above. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: The first part is correct. For the second part (2), it is plausible to believe that the operability of the Ice Condenser Bed may be effected by the door being open. However, the applicant is given information in the stem of the question to indicate that Ice Condenser Bed operability has not been affected. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: First part (1) is correct. For the second part (3). it is plausible if the applicant confuses the Door Monitoring System with the Ice Condenser Door itself. However, the door monitoring system_is_providing indication as required. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: First part (1) is correct. For the second part (2), it is plausible to believe that the operability of the Ice Condenser Bed may be effected by the door being open. However, the applicant is given information in the stem of the question to indicate that Ice Condenser Bed operability has not been affected. e third part (3) is plausible if the applicant confuses the SLC for the door monitoring system with the spec for the doors (Part 1). If so, it is plausible to believe that the SLC is applicable in addition to the TS for the doors. Basis for meeting the KA The applicant is given a condition where an Ice Condenser Door is opened and is asked to determine the effect that this malfunction will have on the Ice Condenser system (i.e. the applicability of Tech Specs and the SLC). Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step. First, the applicant must analyze the conditions given to determine the condition of the ice condenser system. The applicant must then compare that analysis to the Tech Specs! Licensing Commitment listed and determine which of them apply based on the condition of the ice condenser system. -________ Basis for SRO only _-_ _-. , Job Level Cognitive Level , QuestionType , Question Source RO Comprehension BANK 2006 NRC Q42 (Bank 648)

---

Development References Student References Provided rences: TS 3.6.13 and bases TS 3.6.12 SLC 16.6.3

2. OP-MC-CNT-NF Section 2.1.3 SYSO25 K6.OI - Ice Condenser System owledge of the effect of a loss or malfunction of the following will have on the ice condenser system: (CFR: 41.7! 45.7) pper and lower doors of the ice condenser Tuesday, July 13, 2010 Page 43 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION__15 L 2515 401-9 Comments: Remarks!Status 401-9 Comments: No comment. Resolution / Comments: N/A Tuesday, July 13, 2010 Page 44 of 294

Question 15

References:

From Lesson Plan OP-MC-CNT-NF Section 2.1.3: Objective #3 2.1.3 Ice Condenser Doors (refer to Drawing 7.1) The purpose of the lower ice condenser doors during normal plant operation is to:

Provide a flow barrier from lower containment to the ice condenser lower plenum.
Provide thermal insulation around the lower crane wall.

During accident conditions, its purpose is to provide a path into the ice condenser from lower containment on a pressure increase due to a LOCA. There are 24 pairs of inlet doors which have a total flow area of 1000 ft

                                                                      . The doors require a force of 1 2

2 to fully open (refer to Drawing 7.3). The doors will open slightly if the cold air head is removed lb/ft from the ice condenser. Shock assemblies are provided to dissipate the energy of rapid door opening for large break accidents. Objective #2 Each of the lower ice condenser inlet doors have sensors which will generate the alarm Ice Condenser Lower Inlet Doors Open on 1(2)AD-9 in the control room.

ICE COND LOWER INLET DOORS OPEN Setpoint: Door (any of 24) pI fully closed.

Origin: Limit switches monitoring the lower inlet doors. (MC1NPLL-6000, through 1NFLL691O) Probable Cause: 1. Improper operation of the Containment Air Return Fans.

2. Malfunction of Containment Pressure Control System.

3. LOCA

4. Steam Line or Feedwater line break.

Automatic Action: None Immediate Action: 1. Check proper operation of the Containment Air Return Fans AND VQ System.

2. IF alarm was jQ] caused by a high energy line break, send operator to close affected door(s).

3. Refer to Tech Specs.

4. IF malfunction of CPCS, notify WCC SRO.

5. Notify Engineering to evaluate potential water intrusion into the ice condenser floor.

(PIP 2-M97-2686)

From TS 36.13: 3.6 CONTAINMENT SYSTEMS 3.6.13 Ice Condenser Doors LCO 3.6.13 The ice condenser inlet doors, intermediate deck doors, and top deck doors shall be OPERABLE and closed. APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS NOTE - Separate Condition entry is allowed for each ice condenser door. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more ice A.1 Restore door to 1 hour condenser doors OPERABLE status. inoperable due to being physically restrained from opening. B. One or more ice B.1 Verify maximum ice bed Once per 4 hours condenser doors temperature is 27°F. inoperable for reasons other than Condition A or AND not closed. B.2 Restore ice condenser door 14 days to OPERABLE status and closed positions. (continued)

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2006 CNS SRO NRC Examination QUESTION 42 648 N/A - N/A Never Assigned to a K/A Given the following:

Unit 1 is in Mode I at 10% power.
ICE COND LOWER INLET DOORS OPEN alarm is lit.
The lower inlet door position display panel indicates that a door is open.
The door is confirmed to be cracked opened. The door will not move further open and cannot be closed.
No other alarms related to the ice condenser, NE system or AHUs are lit.

Which one describes the Tech Spec and/or SLC that must be entered for the current plant conditions? A. EnterTech Spec 3.6.13 Ice Condenser Doors B. Enter SLC 16.6-3 Ice Condenser Door Position Monitoring System C. Enter Tech Spec 3.6.13 Ice Condenser Doors, and Enter SLC 16.6-3 Ice Condenser Door Position Monitoring System D. Enter Tech Spec 3.6.12 Ice Bed, and Enter Tech Spec 3.6.13 Ice Condenser Doors Wednesday, July 14, 2010 Page 84 of 203

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2006 CNS SRO NRC Examination QUESTION 42 General Discussion Per TS 36.13 (Ice Condenser Doors) The ice condenser inlet doors, intermediate deck doors, and top deck doors shall be OPERABLE and closed while in MODES 1, 2, 3, and 4. Answer A Discussion Answer B Discussion LT door monitoring system is providing indication as required. Answer C Discussion rThe door monitoring system is providing indication as required Answer D Discussion [er information provided in the stem, the ice bed temperature is unaffected by the malfunction. Basis for meeting the KA Basis for Hi Cog LBasis for SRO only E Job Level Cognitive Level QuestionType Question Source RO Memory BANK L Development References Student References Provided eferences:

1. TS 3.6.13 and bases

2. OP-CN-CNT-NF1 1 N/A -N/A Never Assigned to a K/A 401-9 Comments: RemarkslStatus Wednesday, July 14, 2010 Page 85 of 203

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 16 SYSO26 A2.09 Containment Spray System (CSS)

( bility to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to rrect, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5/45.3 /45.13) Radiation hazard potential of BWST Given the following conditions on Unit 1:

The unit is in MODE 4 with A Train ND in service
A CRUD Burst has been initiated and clean up is in progress
Unknown to the Operators, 1 ND-35 (ND Sys to FWST Isol) has developed a small leak past its seat (0.5 GPM)

1. Which of the following describes the operational concern associated with this condition?

2. In accordance with OP/I /A/6200/014 (Refueling Water System) what alignment would be required to address the radiological effects of this event?

A. 1. In creased radiation levels at the FWST enclosure.

2. The FWST would be placed in purification with the FW pump for clean up.

B. 1. Increased radiation levels at the FWST enclosure.

2. The FWST suction piping would be placed in recirculation using the FW Recirc pumps to dilute the crud deposited in the ECCS suction piping.

C. 1. The formation of hot spots in the ECCS suction piping downstream of 1FW-27 (FWSTt0 ND Pump Isol).

2. The FWST would be placed in purification with the FW pump for clean up.

D. 1. The formation of hot spots in the ECCS suction piping downstream of 1FW-27 (FWSTt0 ND Pump Isol).

2. The FWST would be placed in recirculation using the FW Recirc pumps to dilute the crud deposited in the ECCS suction piping.

Tuesday, July 13, 2010 Page 45 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 16 25161 General Discussion In the situation given, the ND system would be highly contaminated due to being aligned for RHR during a crud burst clean up. Leakage past IND-35 would result in the highly contaminated ND system water entering the FWST. This would result in increased radiation levels at the FWST tank inside the FWST enclosure located outside the RCA in yard west of the Unit 1 containment building. To address this issue the U-i FWST pump would be placed in purification using the U-i FW pump and KF demineralizers. Answer A Discussion CORRECT: See explanation above. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Answer 1 is correct. Answer 2 is plausible because the line to the FWST from ND ties in to the ECCS suction piping upstream of 1FW-27 (Normally Open) and it would be conceivable for the applicant to conclude that contamination of the ECCS suction piping would be the operational concern. However this would be incorrect because in an RHR alignment, 1FW-27 is closed and there is no flowpath to the ECCS suction piping. If this was actually a concern, the FW Recirc pumps are designed to recirc the water in this_piping_and therefore this would be aresonable answer. Answer C Discussion liRRECT See explanation above. PLAUSIBLE: Answer I is plausible because the line to the FWST from ND ties in to the ECCS suction piping upstream of 1FW-27 (Normally Open) and it would be conceivable for the applicant to conclude that contamination of the ECCS suction piping would be the operational concern. However this would be incorrect because in an RHR alignment, I FW-27 is closed and there is no flowpath to the ECCS suction piping. Answer 2 is correct. Answer D Discussion - CORRECT: See explanation above. PLAUSIBLE: Answer 1 is plausible because the line to the FWST from ND ties in to the ECCS suction piping upstream of 1FW-27 (Normally Open) and it would be conceivable for the applicant to conclude that contamination of the ECCS suction piping would be the operational concern. However this would be incorrect because in an RHR alignment, IFW-27 is closed and there is no flowpath to the ECCS suction. Answer 2 is plausible is plausible because if answer 1 were correct the FW Recirc pumps are designed to recirc the water in this piping and therefore this would be a reasonable answer. Basis for meeting the KA - -. This K/A is matched because both the FWST and ND (Aux Spray) are integral components of the CSS. The NS system takes suction from the FWST and sprays the water directly to the containment building so it is physically impossible for NS to pose a radiation hazard to the FWST unless it had been aligned for sump recirculation. It this case, the FWST would have been depleted and the need to align it for clean up would not exist. In order to have any plausible scenario where the CSS could pose a radiation hazard to the FWST, the ND system must be the system providing the source for the increased radiation. The scenario described requires the applicant to evaluate a malfunction (Leakage past a normally closed isolation valve) and predict the operational impact. He must then pick the correct strategy to mitigate the consequences of the malfunction. Basis for Hi Cog This is a hi cog question because it involves a level of analysis of the given situation, application of system knowledge and solving a problem related to the effects and how to mitigate those effects. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW veloprnent References -- Student References Provided Lesson Plan OP-MC-FH-FW Pg 17 (Rev 41) L Tuesday, July 13, 2010 Page 46 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 16 25 SYSO26 A2.09 Containment Spray System (CSS)

i1ity to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to rrect, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5/43.5/45.3/45.13) Radiation hazard potential of BWST O1-9 Comments: RemarkslStatus L__ 401-9 Comments: A procedure of reference needs to be in affect to satisfy Part 2 of the KA Suggestion: Add a procedure to part two of the stem: What alignment per procedure XXXX would be required... Resolution / Comments: Added procedure reference to question 2 of the stem. See attached file for proposed revision to question. Tuesday, July 13, 2010 Page 47 of 294

Question 16

References:

From OP-MC-FH-FW Pg 17 (Rev 41)

1.0 INTRODUCTION

1.1 Purpose Objective I The purpose of the Refueling Water System is to provide a source of borated water to be used during refueling, LOCA or as makeup for the Spent Fuel Pool. The system can remove impurities from the refueling cavity and transfer canal during refueling. The system can clean up the FWST. The system also provides a means of transferring the refueling water between the refueling cavity and the FWST. 1.2 General Description The system consists of: . The FWST.

The FW Pump.
The FW Recirculation pumps.
Four 30 kilowatt heaters connected into three 40 kilowatt heater groups.

The borated water is used during refueling to flood the refueling cavity. There is sufficient static head to partially fill the refueling cavity. The FW pump completes the fill and drain process. Water can be passed through the KF demineralizers for clean up during filling and draining. The FWST Recirculation Pumps The Recirculation pumps are used to maintain a 70°F temperature in the suction header. This header supplies suction to the NV, ND, NI, and NS pumps. The suction header volume is recirculated every 2.5 hours.

From OP-MC-FH-FW Pg 17 (Rev 41) When raising or lowering level in the refueling canal, there is a potential for an airborne contamination event to occur. To prevent this from occurring during filling operations, we normally place VP in service and maintain a fill rate low enough to preclude airborne problems. To preclude this problem from occurring during drain down operations, we wash down the cavity walls while draining. Since demin water is used to wash down the walls, there is a potential for a dilution of either the NCS and/or the FWST to occur. To control that problem, we use the enclosure in the FW procedure to calculate and limit the amount of demin water used based on the system boron concentration and the volume of water in the refueling cavity. 23 Refueling Water Cleanup The water in the refueling cavity or FWST can be recirculated any time for cleanup. Recirculation is accomplished using the FW pumps and KF demineralize rs. To cleanup the refueling cavity, the suction of the FW pump is aligned to the Refueling Cavity. The FW pump discharge is then aligned to the KF purification loop. The discharge of the purification loop is then routed back to the refueling cavity. To clean up the FWST, the suction of the FW pump is aligned to the FWST. The discharge of the FW pump is then aligned to the KF purification loop. The discharge of the purification loop is then routed back to the FWST. 2.4 Recirculation of the Refueling Water In order to maintain the suction header to the ECCS pumps greater than 70°F at all times, the water is circulated back to the heated FWST via the FWST Recirculation pumps. 2.5 Design Basis of the Refueling Water System The Refueling Water System is designed to provide:

A source of borated water at refueling water boron concentration for use during refueling or a postulated LOCA.
Recirculate the refueling cavity and transfer canal for cleanup during refueling.
Recirculate the water in the FWST for cleanup following refueling.

LJ D LC INSIDE CONTAINMENT A C 0 B L D OUTSIDE CONTAINMENT L E G S H 0 T B L E C G S Transferring Water from the FWST to the Refueling Cavity using the ND Pump

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 17 SYSO39 K4.05 Main and Reheat Steam System (MRSS) nowledge of MRSS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) ( Automatic isolation of steam line Given the following conditions on Unit 1:

A LOCA has occurred inside Containment
Containment pressure is 3.4 PSIG
The crew is preparing to initiate a cooldown per ES 1.2 (Post LOCA Cooldown and Depressuration)

Which ONE (1) of the following must occur to allow reopening the MSIVs for the given conditions? A. Reset the Main Steam Isolation signal ONLY. B. Reset the Phase B AND Main Steam Isolation signals ONLY. C. Containment pressure must be reduced below 3 PSIG AND reset the Main Steam Isolation signal ONLY. D. Containment pressure must be reduced below 3 PSIG AND reset BOTH the Main Steam Isolation signal and Phase B Isolation signal. Tuesday, July 13, 2010 Page 48 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 17 25171 neraI Discussion In the scenario given, a Main Steam isolation would have occurred due to a containment pressure reaching 3 PSIG. (Hi Hi Containment Pressure) In order for the operator to reset the main steam isolation signal, the crew would only be required to reset the MSI signal. The MSI signal associated with the Hi Hi containment pressure allows reset at any time_regardless of Containment_pressure. - Answer A Discussion - CORRECT: See explanation above. Answer B Discussion - INCORRECT: See explanation above. PLAUSIBLE: A high high containment pressure does exist (containment pressure> 3 psig). This condition would result in a Phase B isolation and a Main Steam isolation. It is plausible that the applicat could misinterpret the actuation of a MSI as being the result of a Phase B isolation. It would therefore be plausible to consider it necessary to reset both the Phase B signal and the main steam isolation signal to reopen the MSIVs. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: A high high containment pressure does exist (containment pressure> 3 psig). Some of the MSI actuation signals such as low S/G pressure require the signal to be cleared or blocked in order to reset MSI. It is plausible that the application would misinterpret Hi Hi containment pressure as being one of those signals and conclude that containment pressure must be reduced to reset the MSI signal. Answer 0 Discussion . . - INCORRECT: See explanation above. PLAUSIBLE: A high high containment pressure does exist (containment pressure> 3 psig). This condition would result in a Phase B isolation and a Main Steam isolation. It is plausible that the application would misinterpret the actuation of a MSI as being the result of a Phase B isolation. It would therefore be plausible to consider it necessary to address this condition prior to resetting the main steam isolation. Additionally, some of the MSI actuation signals such as low S/G pressure require the signal to be cleared or blocked in order to reset MSI. It is plausible that the application would misinterpret Hi Hi containment pressure as being one of those signals and conclude that containment ssure must be reduced to reset the MSI signal. sis for meeting the KA ts./A is matched because the applicant is required to evaluate a given scenario where a main steam isolation has occurred (Auto isolation of steam [_djosess knowledge of the MRSS design features and interlocks to determine actions required to allow reset of this signal. Basis for Hi Cog This is a hi cog question because it involves a level of analysis of given situation, apply system knowledge and solve a problem of what would be required to affect a reset of the main steam isolation signal. .

                                                                                                                                    --             _i Basis for    SRO   only N/A Job Level            Cognitive Level             QuestionType                                        Question Source RO                Comprehension                  BANK                                     MNS Bank Q # ASTMSMRO5 Development References                                                                            Student References Provided             -

OP-MC-STM-SM Rev 25 Pg 33 and 37 OP-MC-STM-SM Learning Objective #10 SYSO39 K4.05 Main and Reheat Steam System (MRSS)

Knowledge of MRSS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) Automatic isolation of steam line 401-9 Comments: RemarkslStatus 10l-9 Comments: In C and D: add be between must and reduced. C is a subset of D. If C was correct, D would be also. C needs an only. Wednesday, July 14, 2010 Page 49 of 294

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2010 MNS SRO NRC Examination QUESTION 17 25l7 Resolution I Comments: Added be in distracters C and D and ONLY at the end of distracter C as recommended by Lead Examiner. See attached file for revised version of question. -- Tuesday, July 13, 2010 Page 50 of 294

Question 17

References:

NNL L L No. OBJECTIVE L L P P 0 OORSR ROO

9. Concerning the SIG PORVs;

    . List the open and close set points.                        X X X X X
    . Discuss local operation of the valves.                     X X X X X
    . Describe the operation and control of the valves.          X X X X X

10. Concerning the Steam line Isolation signals;

    . List the Steam line Isolation signals.                     X X X X X
   . List the components affected by a MSI signal               X X X X X
   . State the set points and logic requirements for initiation. X X X X X
   . Describe any operator actions required to reset the signal.

X X X II. Concerning the Main Steam line Isolation Valves;

   . State the purpose of the Main Steam line Isolation Valves. X X X X
   . Describe the operation and control during opening and           X X X closing.
   . Explain the importance of establishing less than 50 psid        x x x across the MSIVs prior to opening.

12. Evaluate plant parameters to determine any abnormal system X X X conditions that may exist.

13. Concerning the Tech Specs related to the Main Steam System;

   . Given the LCD title, state the LCD (including any COLR          X X X values) and applicability
   . For any LCDs that have action within one hour, state the       x x x action.
   . Given a set of parameter values or system conditions,           x x x determine if any Tech Spec LCD(s) is(are) not met and any action(s) required within one hour.
   . Given a set of plant parameters or system conditions and        x x x the appropriate Tech Specs, determine the required action.
   . Discuss the basis for a given Tech Spec LCD or Safety             X

Limit.

SRO Only

From OP-MC-STM-SM Pg 37 0.1. Instrumentation and Controls Main Steam Isolation Initiation/Reset pushbutton [ RESET INITIATE INITIATE = Indicating Lights RESET Objective #10 Two pushbuttons, Train A and Train B, used to initiate or reset MSI signal Resets MSI signal if signal is cleared or blocked, except the MSI may be reset with the Hi-Hi Containment Pressure signal still present. PORVIMSIV Bypass Reset Pushbuttons Two reset pushbuttons (Train A & Train B) RESET =lndicating Light RESET Allows Reset of PORV andlor MSIV Bypass valves (MSI signal must be reset before PORV or MSIV Bypass can be reset)

From OP-MC-STM-SM Pg 33 IF MSIVs will not close, there are several different procedures that will reference an enclosure for locally closing MSIVs. EPIE-2.1 Uncontrolled Depressurization of All Steam Generators Enclosure 3 will locally close the MSIVs. The direction is to remove power from the solenoids by opening breakers on EVDA breaker 18 and EVDD breaker 23. IF valves still not fully closed then Maintenance will assist OPS in removing control air from the valves. A test circuit for each valve was provided so that the valve could be tested during plant operation. This 90% partial stroke circuit is no longer required to be performed, so therefore a NSM MG-12563 has removed this test and switch. Objective #11 To prevent a Main Steam line Isolation from occurring, due to a rapid decrease in SIG pressure, a differential pressure of less than 50 psid must be established across the MSIVs prior to opening them. MSIVs receive an automatic close signal for any of the following:

Hi-Hi Containment Pressure
Low steam line pressure> P-il
High rate of pressure decrease < P-Il (if low steam line pressure is blocked)
Manual Isolation pushbutton

Parent Question: Question 620 ASTMSMRO5 ASTMSMRO5 1 Pt Given the following conditions:

A LOCA is in progress
Containment Pressure is 4.2 psig Which of the following must occur to allow reopening the MSIVs for the given conditions?

A. The Main Steam Isolation must be reset B. Containment pressure must decrease below 3 psig and the Main Steam Isolation signal must be reset C. Containment pressure must decrease below 3 psig, the Main Steam Isolation signal must be reset and the Phase B Isolation signal must be reset D. The Main Steam Isolation signal must be reset and the Phase B Isolation signal must be reset Answer 620 A STM-SM, objective 10 Section 2.10

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 18 SYSO59 2.4.31 Main Feedwater (MFW) System

YS059 GENERIC

.now1edge of annunciator alarms, indications, or response procedures. (CFR: 41.10 I 45.3)

Given the following conditions on Unit 1:

The unit is operating at 45% RTP
Channel I Main Turbine Impulse pressure indicates 310 PSIG
Channel 2 is indicating 0 PSIG
The AMSAC UNBLOCK light is DARK Which ONE (1) of the following describes the current status of the AMSAC system?

A. Auto actuation is NOT functional. The loss of CF flow path auto actuation can NOT be restored until the failed impulse channel is repaired. B. Auto actuation is NOT functional. The loss of CF flow path auto actuation can be restored by depressing the AMSAC Actuation U N BLOC K pushbutton. C. Auto actuation will occur if both Feedwater pumps trip. The loss of CF flow path auto actuation can NOT be restored until the failed impulse channel is repaired. D. Auto actuation will occur if both Feedwater pumps trip. The loss of CF flow path auto actuation can be restored by depressing the AMSAC Actuation UNBLOCK pushbutton. Tuesday, July 13, 2010 Page 51 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 18 2518 neraI Discussion During power escalation when Turbine impulse pressure reaches 290 psig on 2/2 channels, the circuit will automatically reinstate full AMSAC protection. In the scenario given in this question, one channel has failed low and therefore this auto reinitiation did not occur and the UNBLOCK light is dark. The circuit can be unblocked any time the IJNBLOCK pushbutton is depressed and if at least one turbine impulse channel is greater than 290 psig. the circuit will remain armed. AMSAC has two actuation signals: one is due to a loss of CF flowpath. This feature is blocked and unblocked as described above. The second is due to a loss of both CF pumps and cannot be blocked and therefore always active. - - Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part (1) is plausible if the applicant confuses the block/ unblock function of AMSAC as applicable to both the loss of CF flowpath and the loss of CF pumps. Part (2) is plausible if the applicant confuses the requirement for 2 channels >290 psig as being required in order to either automatically or manually reinstate AMSAC protection. This would seem reasonable as a fail safe mode for the circuit. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Part (1) is plausible if the applicant confuses the block/ unblock function of AMSAC as applicable to both the loss of CF flowpath and the loss of CF pumps. Part (2) is correct and therefore plausible. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: Part (1) is correct and therefore plausible. t (2) is plausible if the applicant confuses the requirement for 2 channels >290 psig as being required in order to either automatically or

      .iually reinstate AMSAC protection. This would seem reasonable as a fail safe mode for the circuit.

Answer D Discussion - -

CORRECT: See explanation above. Basis for meeting the KA K/A is matched because the applicant is required to evaluate a given set of indications associated with the AMSAC system and apply system knowledge to determine the current status. Basis for Hi Cog -- The applicant is required to evaluate a given set of indications and apply system knowledge associated with the MFW system to predict an outcome and solve a problem concerning how arming the system is either possible or not possible in the present situation. Basis for SRO only - Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK MNS Bank Q # CFCFO44 Development References Student References Provided Lesson Plan OP-MC-CF-CF Rev 34. Pgs 41-43 -_____________ OP-MC-CF-CF, Objective 16 SYSO59 2.4.31 - Main Feedwater (MFW) System

/ S059 GENERIC

(\ wledge of annunciator alarms, indications, or response procedures. (CFR: 41.10 / 45.3) 401-9 Comments: RemarkslStatus L 40 1-9 Comments: Wednesday, July 14, 2010 Page 52 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination 8 QUES L TION_ r _l In distractor C: add the before Loss. Change cannot to can NOT Resolution I Comments: Changed question per Lead Examiners recommendation. Distracter A need a the before Loss also. See attached file for revised version of question. Tuesday, July 13, 2010 Page 53 of 294

Question 18

References:

s NNLLL E OBJECTIVE L L P P 0 OORSR R00 11 Explain why and how CF flow is transferred from the CF X X X X nozzle to the CA nozzle and vise versa. 12 Explain the Feedwater Isolation actuation circuit. X X X X X 13 List the CF valves that isolate on a Feedwater Isolation xxxx x Signal. 14 Describe the automatic actions that occurs on:

Hi Hi Doghouse Level
Hi Hi SIG Level (P-14).

15 Explain the purpose of the Anticipated Transient Without Scram Mitigation System Actuation Circuitry (AMSAC). 16 Concerning AMSAC:

State the automatic action that occurs as a result of an AMSAC signal
List the parameters that will actuate the AMSAC automatic actions
Discuss the development of the actuation signals, to include the components monitored and setpoints.

17 Concerning the AMSAC Block/Unblock switch:

State the purpose of each position (Block and Unblock)
State when each position may be used.
Differentiate between the Manual block function and Auto block functions (include the 2 minute time delay for auto blocking).

From OP-MC-CF-CF Pg 41 2.11 ATWS Mitigation System Actuation Circuitry (AMSAC) I Objective #15 The purpose of this circuit is to:

Prevent NCS overpressurization during a loss of main feedwater accompanied by an Anticipated Transient Without a Scram (ATWS).
Trip the Main Turbine and start both motor driven CA pumps if a loss of main feedwater occurs or is anticipated.

Objective #16, 17, 18, 19, 20, 21 The AMSAC circuit monitors conditions that are indicative of an ATWS event as required by 1 0CFR5O.62. An AMSAC actuation decreases the severity of an ATVVS event by minimizing the peak pressure in the NC System. When actuated, AMSAC will:

Trip the Main Turbine
Start both motor driven CA pumps Light the AMSAC Turb Trip Annunator (IADI-BI)

ThAMSAC: circuit can bejvided into three sections: n

Loss of CF flow thtotheSIGs logic r Block/Unblock 1c The loss of both CF pump logic for AMSAC uses six pressure switches (three for each CF pump) which monitor each pumps control oil pressure (refer to Drawing 7.13).

Note: The loss of both CF pumps logic for AMSAC is different than the normal scheme that senses loss of both CF pumps. The normal scheme only uses two (2) pressure switch per pump (1LPPS5180 and 1LPP55184 for A CF pump and 1 LPPS51 90 and 1 LPPS51 94 for B CP pump). Control oil pressure is used to hold the CF pump turbine stop valves open. If 2 out of 3 pressure switches on the same CF pump turbine sense a loss of oil pressure (setpoint < 45 psig) a signal will be sent to AMSAC. If both CF pumps lose control oil pressure, AMSAC will be actuated on a loss of both CF pumps. s obh CF pu MSACauationcannotET3c A selector switch at each CF pump local control panel provides a means for defeating each pressure switch (one at a time) for testing.

From OP-MC-CF-CF Pg 43 The loss of CF flow path logic looks at the CF containment isolation valves, SIG CF control valves (main control valves and bypass valves), and combinations of these which could result in a loss of CF flow to the SIG5. (refer to Drawing 7.14)

When a CF containment isolation valve goes fully closed a signal is sent to AMSAC. If 3 out of 4 CF containment isolation valves close, AMSAC will actuate (if unblocked).
A CF main feed regulating valve (FRV) must be at least 25% open if its associated bypass valve is not fully open in order for AMSAC to consider the flow path viable. Each SIG has a status light AMSAC SIG A(B)(C)(D) Low Flow on I S14-A4(B4)(C4)(D4) to warn the operator of the condition of the flow path for that SIG. If the flow path for one SIG does not meet these criteria for 30 seconds, a status light Any CF SIG path closed > 30 sec on ISI4-D3 will warn the operator. If 3 out of 4 flow paths are not viable for 30 seconds, AMSAC will actuate. The 30 second time delay allows operator action or for the transients to stabilize. Annunciator 3 of 4 CF SIG Paths Closed > 30 sec on IAD4, F5 indicates that AMSAC will be actuating (assuming it is not blocked).
Any combination of containment isolation valve closure or CF main FRV closure from 3 out of 4 SIG flow paths will actuate AMSAC.

.ockIUnblock logic only applies to Loss of(,r, ... path logic. tt.a loss of both CF pumps logic is always active. When tripping the last CF pump during plant shutdown, the operator can depress the CA motor driven pump auto start defeat switch to prevent the auto start of the CA pumps. AMSAC ACTUATION BLOCK1UNBLOCK INDICATING LIGHT UNBLOCK PUSHBUTTONS UNBLOCK

Parent Question CFCFO44 1 Pt Given the following conditions on Unit 1:

           -  Unit I is at approximately 45% power.
           -  One channel of Main Turbine Impulse pressure indicates 310 psig while the other channel indicates 0 psig.
           -  The AMSAC UNBLOCK light is dark Which ONE of the following describes the current status of the AMSAC system?

A. No AMSAC system auto actuations are currently functional, but can be returned to service by depressing the AMSAC Actuation UNBLOCK pushbutton. B. No AMSAC system auto actuations are currently functional, but can be returned to service by depressing the AMSAC Actuation BLOCK pushbutton. C. An automatic AMSAC actuation will occur if both Feedwater pumps trip. The Loss of CF flow path auto actuation can be restored by depressing the AMSAC Actuation UNBLOCK pushbutton. D. An automatic AMSAC actuation will occur if both Feedwater pumps trip. The Loss of CF flow path auto actuation can not be restored until the failed impulse channel is repaired. Answer 61 C

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 19 SYSO61 K2.O1 Auxiliary / Emergency Feedwater (AFW) System

nowledge of bus power supplies to the following: (CFR: 41.7) . AFW system MOVs Given the following conditions on Unit 1:

The unit has experienced a Loss of all AC
Crew has implemented ECA 0.0 (Loss Of All AC Power)
Unit I Control has been swapped to the SSF
An NEC has been dispatched to close the feeder breaker for 1CA-161C (CA Suction Hdr RN Supply Isol)

Based on the conditions described above which ONE (1) of the following states where the NEC would be dispatched to perform this action? A. SMXG B. SMXG1 C. SDSP-1 D. 1EVDA-1 Tuesday, July 13, 2010 Page 54 of 294

__ ___ _____ ________ _____ _____ __ FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 19

  .jeneral Discussion SDSP-l is the power supply to 1CA-161C. This is a 250V power panel board located in the SSF and remains energized when contro transferred to the SSF.                                                                                                               l is Answer A Discussion INCORRECT: See explanation above.

PLAUSIBLE: MCC SMXG is located at the SSF and would be powered by the SSF DIG after a transfer of control to SSF related valves are powered from this MCC but not 1CA-161. the SSF. A number of Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: MCC SMXGI is located at the SSF and would be powered by the SSF D/G after a transfer of control to SSF related valves are powered_from this MCC but not ICA-161. the SSF. A number of Answer C Discussion CORRECT: See explanation above. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: Power panel board 1EVDA-l is the supply to 1NC-2 72C and 1NC-273C which are train C valves controlled from would reasonable for the applicant to include 1CA-161C with this power the SSF. It supply. Basis for_meeting the KA This K/A is matched because in order to correctly answer the questio n the applicant must utilize knowledge of the bus power supply l6lCwhichisaAFWsystemMOV. to 1 CA

Basis_for Hi Cog

             -z isis for SRO only   -

Job Level Cognitive Level QuestionType Tj estion sou__Zj RO Memory NEW Development References Student References Provided Lesson Plan OP-MC-CF-CA Rev 43 Pg 23 SYSO6I K2.0 1 Auxiliary I Emergency Feedwater (AFW) System

Knowledge of bus power supplies to the following: (CFR: 41.7) AFW system MOVs 401-9 Comments: RemarkslStatus 401-9 Comments: No comment. Resolution I Comments: Tuesday, July 13, 2010 Page 55 of 294

Question 19

References:

From OP-MC-CF-CA Pg 23 (Rev 43) In the event of a seismically induced failure of the CAST, no accident mitigation system or safety-related equipment will be adversely affected by flooding of the plant yard. The plant storm drainage system and other design features will prevent flooding of any safety-related structures, systems or components. The flooding of the Turbine Building from a failure of the CAST could release 300,000 gallons (40,000 cubic feet) of water into the Turbine Building. The water flows from penetrations in the mezzanine floor of the stairwells and migrates to the Turbine Building Sumps. The volume of the sumps and pits below the Turbine Building basement elevation is 72,100 cubic feet. Since the volume of the water in the tank is less than the sump volume, the sumps would contain this volume. 2.3.2 Assured Suction Source Objective#1O,11,12 Nuclear Service Water (RN). RN is the safety related water source for the CA system. The RN suction source will align automatically on low CA pump suction pressure (3.5 psig for 2.5+/-.5 sec, except 2A pump is 4.5 psig). The supply valves, RN-i 62B, CA-i 8B and CA-il 6B, are NORMALLY CLOSED. The DG HX inlet valves on B train are interlocked such RN-171B will open when CA-18B opens. The DG interlock ensures that adequate RN flow is available to supply the CA pumps. This only applies to B train. The A train RN supply to CA is on the inlet to the DG HX and has a higher supply pressure and does not go thru the KD HX to the CA suction. NOTE: Automatic cycling the RN supply to CA suction valves is considered an ESF Actuation and is reportable per RP-10 unless intentionally cycled for maintenance. Standby Shutdown Facility Supply (RN/RC). Supply valves CA-161C and CA 162C will automatically align to the Turbine Driven CA pump for SSF operation (3 psig for 2.5+/-.5 sec) and are also opened from the SSF when it is activated for an event. The RN system provides the flow path from the embedded condenser circulating water pipe to provide this assured source. NOTE: 1CA-i6iC and 1CA-162C are DC powered valves from SDSP lD and 2C. On low pressure signal, control power locks-in signal until reset by fully opening valves iCA-i6iC and 1CA-162C. Pulling the DC power fuses will also reset signal.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 20 252 SYSO62 A3.05 AC Electrical Distribution System

                   -
 \bility to monitor automatic operation of the ac distribution system, including: (CFR: 41.7/45.5)

Safety-related indicators and controls Given the following conditions on Unit 1:

B Train of essential equipment is in operation on Unit I
OP/i /A/6350/002 (Diesel Generator) is in progress with the 1A Diesel running in parallel to the grid when the following sequence of events occurs:

o Load is reduced on the diesel to 200KW in anticipation of opening the Emergency Breaker o The RO accidentally OPENS the Normal Feeder Breaker from 1ATC The Blackout Sequencer Actuated Train A status light on Sl-14 Which ONE (1) of the following completes the statement above? A. illuminates and the 1A DG load increases B. remains dark and the 1A DG load increases C. illuminates and the 1A DG Emergency Breaker trips open, then re-closes 8.5 seconds later D. remains dark and the iA DG Emergency Breaker trips open, then re-closes 8.5 seconds later Tuesday, July 13, 2010 Page 56 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 20 2520 neral Discussion Since the Emergency Breaker is already closed no loss of voltage to 1ETA so the Diesel picks up the remaining load on IETA. Answer A Discussion INCORRECT. See explanation above. PLAUSIBLE: This answer is plausible if the applicant does not understand what causes the Sequencer Status Light to actuate. The second part regard ing the DG load increasing is correct. - Answer B Discussion CORRECT. See explanation above. Answer C Discussion - INCORRECT. See explanation above. PLAUSIBLE: This answer is plausible if the applicant concludes that the DG load sheds and then picks the bus back up when the sequencer operates. However, no Blackout relay actuates. It is also plausible if the applicant concludes that the Emergency Breaker trips open on overcurrent. However, the Emergency Breaker will NOT trip on overcurrent if the Essential Bus is disconnected from another power source. Answer 0 Discussion - INCORRECT. See explanation above. PLAUSIBLE: This answer is plausible if the applicant concludes that the DG load sheds and then picks the bus back up when the sequencer operates and does not understand what causes the Sequencer Status light to actuate. However, no Blackout relay actuates. It is also plausible if the applicant concludes that the Emergency Breaker trips open on overcurrent. However, the Emergency Breaker will NOT trip on overcurrent if the Essential Bus is disconnected from another power source. - - Basis for meeting the KA - The KA is matched because the applicant demonstrates the ability to monitor automatic operation of the AC distribution system (i.e. diesel loading) based on DO KW indication and Sequence operation (or in this case it does not operate). Thsis for Hi Cog is a higher cognitive level question because it requires a level of analysis beyond simple memorization. oasisforSROonly . - -- Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 2008 MNS NRC Q21 Development References Student References Provided Lesson Plan OP-MC-DG-EQB Section 2.4 and 3.3 SYSO62 A3.05 AC Electrical Distribution System

Ability to monitor automatic operation of the ac distribution system, including: (CFR: 41.7/45.5) Safety-related indicators and controls 401-9 Comments: RemarkslStatus Proposed replacement for 2010 NRC Q-20. Replacement question appoved RFA 07/06/10 Wednesday, July 14, 2010 Page 57 of 294

Question 20 Proposed Replacement

References:

From DG-EQB Section 2.4: There are an additional 37 lights on the sequencer panel which are used during testing to monitor the signals sent to the 4160V switchgear and 600V load centers. In addition, these lights verify the status of the contacts after testing and serve as the overlapping requirement for switchgear operation. Annunciator alarms in Control Room

Sequencer A (B) in Test, AD-li, B-i (E-1)

This alarm alerts the operator that the sequencer test key switch is in a position other than OFF. The operator has no control of components on the respective bus when this alarm is actuated.

Sequencer A (B) Loss of Control Power, AD-il, B-2 (E-2)

This alarm alerts the operator that the sequencer has lost control power and it will not operate.

   . ETA (ETB) Degraded Voltage, AD-i i, L-i ([

(< 3689 Volts on ETA (ETB) for 10 seconds (88% normal bus voltage)

i. IF a degraded bus voltage condition continues to exist following a i 0 second time delay and a Safety injection signal occurs, a separation o offsite power will be initiated immediately (10 minute time delay is bypassed IF degraded bus voltage condition is not cleared before iO minute elay, the bus will be separated from the offsite power supply by the jpping and lockout of 1 ETA normal and alternate incoming breakers.

3. Loss of voltage (Blackout) sequencer logic will start as a result of trippin he normal and alternate circuit breake Status indication in Control Room (on SI-14)

LOCA Sequencer Actuated Train A or Train B Blackout Sequencer Actuated Train A or Train B
ETA (ETB) Undervoltage on Phase X, Y or Z
ETA (ETB) Degraded Voltage Relays Computer Digital Inputs
Sequencer A (B) Reset complete/not complete
Sequencer A (B) complete/not complete
Sequencer A (B) Control Voltage low/normal
Sequencer A (B) Test Relays actuated/not actuated
Sequencer A (B) LOCA actuated /not actuated
Sequencer A (B) Reset actuated/not actuated There is local indication, a control room annunciator, and a computer point to indicate a loss of control voltage for the 4160-Volt Degraded Voltage Instrumentation

Question 20 Proposed Replacement References Section 2.4: Should a Safety Injection signal occur at any time after the first time delay relay completes its cycle, the circuit will automatically initiate separation from the offsite power source and transfer to the emergency diesel generators. Protection for a severe diesel-generator overload accompanied by a system voltage dip caused by events such as a Loss of Off-Site Power (LOOP) with the Diesel Generator operating in parallel with the grid is provided by the voltage-controlled overcurrent relay (51V). This relay consists of three single phase relays (51VX, 51VY, and 51VZ). The operation of any one of these phase relays will activate an annunciator alarm in the control room (AD-il A-4 (D-4), DIG A (B) Overcurrent) to warn the operator of an overload condition (800 amps @ 3360 volts). Operation of any two of these overcurrent relays will result in operation of the diesel-generator lockout relay (86D). Diesel-generator lockout relay (86D) will trip and lockout the diesel-generator switchgear breaker and initiate a shutdown of the diesel-generator. This lockout must be reset by hand before the breaker can be reclosed. 3.3 Sequencer operation during a Blackout Objective # 5 Sequencer operation during a Blackout with no safety injection signal and the under-voltage is not due to fault relay 86N, 86S or 86B. If 2/3 LOV Relays sense a loss of voltage on their associated 41 60V bus, the blackout relay will pick up and actuate a DIG start. If the UV condition still exists after 8.5 seconds, the blackout logic is sealed in. All 4160V breakers on the bus are then tripped open. When DIG speed is 95%, the output breaker will close. When bus voltage is 92.5% DIG speed is 97%, the accelerated sequence is enabled. Blackout loads will be sequentially applied at intervals of approximately 2 seconds, as long as bus voltage remains 92.5% and frequency remains > 58.2 Hz. Complete loading of all blackout loads, via the accelerated sequence, could be done in as little as 25 seconds. If during the sequencing of blackout loads the Sequencer RESET pushbuttons are depressed, no additional sequencing will occur. This is because once the RESET pushbuttons are depressed, the blackout signal is removed and since there is power on the 4160V bus a blackout no longer exists. It would require another blackout signal or manual loading of the bus to complete the sequencing of loads. Should the Accelerated Sequence Relay scheme fail to work, the Committed Sequence would be actuated approximately 10 seconds after the diesel receives its blackout start signal if load shed of the bus has been completed. The committed sequence may take up to 12 minutes to load all blackout loads. The committed sequence does not require any minimum voltage or minimum frequency to allow it to progress as does the Accelerated Sequence. The Committed Sequence is required by Technical Specifications.

Parent Question 2008 NRC Exam Question 21: Examination Outline Cross- Level RO SRO reference: Tier# 2 Group# 1 KIA# 062A3.05 Importance Rating 3.5 (Ability to monitor automatic operation of the AC distribution system, including: Safety-related indicators and controls) Proposed Question: Common 21 The B Train of essential equipment is in operation on Unit 1. While performing 0P111A163501002 (Diesel Generator) with the 1A Diesel running in parallel to the grid, the following sequence of events occurs:

Load is reduced on the diesel to 200KW in anticipation of opening the Emergency Breaker.
The RO accidentally OPENS the Normal Feeder Breaker from 1ATC.

Which ONE (1) of the following completes the statement below? The Blackout Sequencer Actuated Train A status light on Sl-14 A. lights and the Emergency Breaker trips open, then re-closes 8.5 seconds later. B. remains dark and the Emergency Breaker trips open, then re-closes 8.5 seconds later. C. lights and the 1A DG Load increases. D. remains dark and the 1A DG Load increases.

Proposed Answer: 0 Explanation (Optional): A. Incorrect. Section 3.3 of OP-MC-DG-EQB states that If 2/3 LOV Relays sense a loss of voltage on their associated 41 60V bus, the blackout relay will pick up and actuate a DIG start. If the UV condition still exists after 8.5 seconds, the blackout logic is sealed in. However, with the Emergency Breaker closed, there is NO loss of voltage on I ETA. B. Incorrect. While it is true that the BC signal will not exist, the Diesel Breaker will NOT trip open (Section 2.7 of OP-MC-DG-DG) on overcurrent with the Essential Bus disconnected from another power source. C. Incorrect. No BC Relay will operate (see A) D. Correct. The Emergency Breaker is closed and the Diesel will pick up the remaining load on 1ETA. Technical Reference(s) OP-MC-DG-EQB pages 17, (Attach if not previously 23, 3 provided) OP-MC-DG-DG page 35, (Including version or revision Rev 25 #) Proposed references to be provided to applicants during None examination: Learning Objective: DG-EQB #5 (As available) Question Source: Bank # Modified Bank # McGuire (Note changes or attach NRC Bank parent)

                                                # 187 New Question History:           Last NRC Exam        NA Question Cognitive          Memory or Fundamental Knowledge Level:

Comprehension or Analysis X 10 CFR Part 55 55.41 7 Content: 55.43

Comments: KA is matched because Sequencer indication (status light) is evaluated for a condition where AC Power is interrupted RFA Concurs 4/18108

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 21 SYSO63 A3.O1 - DC Electrical Distribution System bility to monitor automatic operation of the DC electrical system, including: (CFR: 41.7/45.5) Ieters, annunciators, dials, recorders, and indicating lights Given the following plant conditions:

Both Units operating at 100% RTP
A complete Loss of Offsite Power occurred on Unit I
1A DIG started but subsequently tripped on Low Lube Oil Pressure
30 seconds have passed since the Loss of Offsite Power occurred Which ONE (1) of the following describes the condition of the components listed below?

1. 125 VDC Vital Distribution Center (EVDA)

2. Annunciator lAD-i I I B2 (Seq A Loss of Control Pwr)

A. 1. Energized

2. LIT B. 1. Energized

2. DARK C. 1. De-energized

2. LIT D. 1. De-energized

2. DARK Tuesday, July 13, 2010 Page 58 of 294

_______ ______ ______ ________ _____ FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 21 2521

 ..eneral Discussion During a blackout or LOOP event, on one or both trains, the essential motor control centers feeding the vital I & C battery chargers, associated with the affected train, will be load-shed by the diesel generator loading sequencer. Normally the battery chargers would be reloaded but in the scenario given the associated DIG has tripped and is not available. During the time period that the battery chargers are de-energized, the batteries, alone, feed the vital instrumentation and control loads. In this case it would be Battery EVCA feeding power panel board 1EVDA.

Annunciator lAD-li 82 (Seq A Loss of Control Pwr) is fed from 1EVDA. If IEVDA were de-energized this alarm would be Lit. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: First part is correct. Second part is plausible if the applicant does not understand what will cause power to be lost to the sequencer or concludes that on a LOOP control power is lost to cause_sequencer actuation. -________ Answer B Discussion CORRECT: See explanation above. Answer C Discussion -

liRRECT: See explanation above. PLAUSIBLE: Part 1 is plausible if the applicant confuses the loss of the battery charger (caused by the LOOP) with the loss of bus EVDA. If that were the case, the applicant could conclude that EVDA is de-energized. While the battery chargers will have been lost and not supplying the batteries until they are restarted, the battery will continue to supply EVDA until battery voltage decreases to the point that EVDA must be separated from the battery (by procedure). Part 2 is plausible (because it would be correct) if the applicant concludes that EVDA is deenergized. Answer D Discussion - INCORRECT: See explanation above.

   .AUSIBLE: Part 1 is plausible if the applicant confuses the loss of the battery charger (caused by the LOOP) with the loss of bus EVDA. If that were the case, the applicant could conclude that EVDA is de-energized. While the battery chargers will have been lost and not supplying the batteries until they are restarted. the battery will continue to supply EVDA until battery voltage decreases to the point that EVDA must be separated from the battery (by procedure).

Second part of the question correct. is -____________ Basis for meeting the KA This system KIA is associated with the ability to monitor the automatic operation of the DC Distribution system including annunciators. In this question the applicant must understand how the vital DC distribution will operate in a situation where a loss of AC has occurred with the train ally supplying power. He is also asked to evaluate the status of this distribution system based on annunciator indication. Basis for Hi Cog --__________________ This is a hi cog question because it involves a level of analysis of given situation, apply system knowledge and solve a problem of what both would be the effect and how the system would respond to the conditions given in the stem. - Basis for SRO only Job Level[ Cognitive Level QuestionTypJ Question Source RO Comprehension NEW Development References Student Lesson Plan OP-MC-EL-EPL Pg: 37 (Rev 23) OP-MC-EL-EPLObj: 20 -_________

  /S063 A3.01 DC Electrical Distribution System
                    -

Ability to monitor automatic operation of the DC electrical system, including: (CFR: 41.7 145.5) Meters, annunciators, dials, recorders, and indicating lights Tuesday, July 13, 2010 Page 59 of 294

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2010 MNS SRO NRC Examination QUESTION [401 -9 Comments: Remarks!Status E 2 1 40 1-9 Comments: C and D are NP. I disagree with distractor analysis for CI and Dl. In order to give Cl and Dl plausibility, you will have to significantly increase the loss of AC down time to almost battery exhaustion (within minutes). This Q is U because C and D are NP Resolution / Comments: Believe that C and D are plausible. We had one validator that picked C. When asked why they picked C they commented that they had mistakenly concluded that 1 was asking for the status of the 125VDC Battery Charger instead of the Distribution jnter. Tuesday, July 13, 2010 Page 60 of 294

Question 21

References:

18 Describe the function of the following inverter indications and X X X X X controls:

output amps
output voltage
output frequency
alternate AC source input frequency
pre-charge push-button
pre-charge light
in-sync light
alternate source off-frequency light
inverter supplying load light
alternate AC source supplying load light
main semi-conductor fuse failure light
low DC input voltage light
low AC output voltage light
high AC output voltage light
low alternate AC source voltage light
inverter failure light
overtemperature light
alarm bypass circuit trouble light
fan failure light
alarms bypassed key switch 19 Given a Limit and/or Precaution, associated with the 125 VDC X X X X X and 120 VAC Vital Instrumentation and Control Power Systems, discuss its basis and applicability.

20 Describe the expected operation of the 125 VDC and 120 X X X X X VAC Vital Instrumentation and Control Power Systems during a Blackout or LOOP (Loss of Off-Site Power) Event.

3.2 Abnormal and Emergency Operation Objective # 20 During a blackout or LOOP event, on one or both trains, the essential motor control centers feeding the vital I & C battery chargers, associated with the affected train, will be load-shed by the diesel generator loading sequencer. Within eleven seconds after the diesel generator start signal the affected essential motor control centers and battery chargers will be reloaded onto the essential bus by the diesel generator loading sequencer. During the time period that the battery chargers are de-energized, the batteries, alone, feed the vital instrumentation and control loads. Protective diodes, within each battery charger, prevent the associated battery from discharging through its battery charger when the charger is de-energized. Objective # 21 During a safety injection the vital I & C battery chargers are treated as a safety injection load and should remain energized. However, during a blackout condition or a safety injection with a blackout the vital I & C battery chargers will first be load-shed and then reloaded, within eleven seconds, after the diesel generator start signal. API1(2)1A15500107, Loss of Electrical Power directs the operator to realign the vital battery chargers once normal power is available. This is done by:

1. Determining which battery chargers are actually being powered from Unit 1 or Unit 2 (dependent upon which unit experienced the loss of electrical power).

2. Depressing the STOP push-button on the vital battery chargers that are being powered from the opposite unit (Unit 2 if the event was on Unit 1 I Unit I if the event was on Unit 2)

This is done because the loading sequencer will close the rn contacts for all the battery chargers. However, the battery charger will only be receiving power, from the selected MCC, based on the breaker closed within the charger connection box (protected by the Kirk Key Interlock). API1(2)1A1550011 5, Loss of Vital or Aux Control Power provides direction to the operator in diagnosing and responding to a loss of a Vital DC or AC Bus. Refer to current copy of this AP for Symptoms and Immediate Actions. Objective # 23 AP-1 5, Enclosure 1 (Response to Degraded DC Bus Voltage) directs the operator to remove the battery from service if its voltage decreases to 105 Volts. The reason for this is that as voltage decreases, current will increase. This is explained by the formula P=IE (Power = Amps x Volts). The increase in current could damage the loads supplied by the battery. Additionally, the operator should recognize that as the battery voltage decreases and the current increases, the battery discharge will increase steadily until the battery is exhausted.

OP/IfA61OO/OtO L Annunciator Response For Panel lAD-il Page 21 of 153 Nomenclature: SEQ A LOSS OF Window: CONTROL PWR B2 Setpoint: NA Origin: Alarm relay (74A) on Sequencer A control power Probable Cause: Feeder breaker 1EVDA-6 open Automatic Action: Sequencer inoperable Immediate Action: L cause of alarm is known (expected alarni) AND Sequencer logged in Tech Specs, no fintlier action required.

2. Check closed IEVDA-6.

Supplementary Action: L Notify SRO. I Refer to Tech Specs.

References:

UFSAR Figure 8-35
MC-1705-O1
MCS-1 14.OO-EQB-0001
MC-1753-Ol.Ol
MC-l 765-00M2
MCEE-1 14-00XJ6-Ol
Tech Specs End Of Response Unit 1

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 22 L SYSO63 K4.Ol DC Electrical Distribution System

nowledge of DC electrical system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) v1anual/autornatic transfers of control Given the following plant conditions:

Both Units are operating at 100% RTP
Battery 1DP is aligned for equalizing charge
The DC Output breaker for Charger 1 DS has tripped open Which ONE (1) of the following describes the current status of Bus IDP?

A. Bus 1 DP will be de-energized. B. Bus I DP will be energized from Charger 1 DP ONLY. C. Bus IDP will be energized from Chargers 1DP and 2DP. D. Bus IDP will be energized from Charger 1DP and Battery 1DP. Tuesday, July 13, 2010 Page 61 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 22 2522

 .eneral Discussion During an equalizing charge, the battery (1DP) is disconnected from its distribution center and the standby charger (1DS) performs the charging operation while the normal battery charger supplies the distribution center. Both distribution centers are electrically cross-connected through the bus tie breakers (Normally Closed). This alignment ensures that there is a battery to supply both busses so that a loss of power will not result in the loss of either bus (I DP or 2DP).

During the battery charge. the DC bus (distribution center) must be disconnected from the battery (1DP) and the charger performing the charge (1 DS) due to the high voltage (approx. 271 VDC) and current conditions existing during the charge. In this alignment, if the charger IDS output breaker trips, the U-I and U-2 250 VDC buses (1DP and 2DP) would be supplied by their normal chargers (1DP and 2DP). On a complete loss of power, buses 1DP and 2DP would be supplied by the 2DP battery. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: The situation described in the stem would result in the loss of battery charger IDS and effectively the loss of battery 1DP as well because both the battery and the standby charger are electrically disconnected from the bus. If the applicant does not remember that during a equalize charge the cross ties between the two unit 250 VDC buses are closed, this could seem a reasonable answer. Also, the applicant could confuse the alignment during an equalize charge with how it is accomplished for the I25V Vital Buses where the Normal charger is used for the equalize charge and the stby charger supplies the bus. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Charger IDP is electrically aligned to bus and is supplying power to it. If the applicant does not remember that during a equalize charge the cross ties between the two unit 250 VDC buses are closed, this could seem a reasonable answer. This information is not given in the stem and therefore required the recall of system knowledge to eliminate it as a possible answer. Answer C Discussion ORRECT: See explanation above. nswer 0 Discussion INCORRECT: See explanation above. PLAUSIBLE: Charger 1DP is electrically aligned to bus and is supplying power to it. If the applicant does not remember that during a equalize charge the battery and the Stby Charger are disconnected from the bus this would be a reasonable choice. All the applicant is given is that the Stby charger output breaker has tripped open, there is no information given about how the system is aligned other than the type of charge that is being performed. . Basis for meeting the KA K/A is matched because the applicant must understand how the 250 VDC system is designed to function during an equalization charge on a battery. The Auto/Manual transfer of control for a DC system is a difficult concept to test because the transfer is passive and dependent on the alignment. This question examines the understanding of how the system will respond to an interruption (loss of charger) and therefore requires_an_understanding of how this passive transfer will take place. Basis for Hi Cog This is a high cog question because it requires more than one mental step. First, the applicant must recall from memory the electrical lineup during and equalizing charge on the 1DP Battery. The applicant must then determine the effect of the DC Output breaker on the battery charger opening on the 1DP Bus. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK MNS Bank Q ELEPJOO7 Development References dent References Provided sson Plan OP-MC-EL-EPJ Pg 17 (Rev 12) OP-MC-EL-EPJ Obj: 5 Tuesday, July 13, 2010 Page 62 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 22 2522 SYSO63 K4.Ol DC Electrical Distribution System

( ow1edge of DC electrical system design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7)

 .Aanual/automatic transfers of control 401-9 Comments:                                                               marks!Status
                                      -

j 401-9 Comments: No comment. Resolution / Comments: L__ Tuesday, July 13, 2010 Page 63 of 294

Question 22

References:

NNLLL OBJECTIVE L L P P 0 OORS R ROO 1 State the purpose of the 250 VDC Auxiliary Power X X X X System. 2 Draw a simplified composite of the 250 VDC X X X X Auxiliary Power System (breaker numbers not required) as provided by Training Drawing 7.2, Simplified Drawing-250 VDC Auxiliary Power System. 3 State the duty-cycle requirements associated with X X X X X the batteries for the 250 VDC Auxiliary Power System 4 List the typical loads powered by the 250 VDC X X X X Auxiliary Power System. 5 Describe the basic operation of the 250 VDC X X X X Auxiliary Power System. 6 Given a Limit and/or Precaution, associated with the X X X X X 250 VDC Auxiliary Power System, discuss its basis and applicability.

From Lesson Plan OP-MC-EL-EPJ Pg 17 (Rev 12) During an equalizing charge, the battery is disconnected from its distribution center and the standby charger performs the charging operation while the normal battery charger supplies the distribution center. Both distribution centers are electrically cross-connected through the bus tie breakers. During this alignment the DC bus (distribution center) must be disconnected due to the high voltage (approx. 271 VDC) and current conditions existing during the charge. The appropriate charging time is determined by the battery condition and the finishing charge parameters. 1(2)DS is equipped with an internal equalize timer which is set by IAE. This timer will automatically place 1(2)05 back to Float when the timer times out. Once the battery is fully charged, the charger is realigned for normal operation with the standby charger disconnected and the bus tie breakers opened. Objective # 6 0.0.1. Limits and Precautions The DC ties will normally remain open. Close only during equalizing charges to batteries, or on loss of battery. Basis: This ensures the DC channels remain independent of each other and that a fault on one bus (channel) does not affect the other bus (channel). Do not allow smoking, open flames, or sparks in the battery area. Basis: Limits the possibility of a fire I explosion due to interaction with hydrogen (a by-product of the electrolytic process of the battery) which could result in damage to equipment vital to the safe operation of the plant. Ensure Battery Room Ventilation Equipment is in service. Basis: Limits the possibility of hydrogen building up to potentially explosive or burnable mixtures which could result in damage to equipment vital to the safe operation of the plant. Distribution Center 1 DP and 2DP are the emergency source of power for critical plant equipment and lighting needed for plant shutdown on loss of offsite power (LOOP). De energizing these Distribution Centers removes that emergency power source. Basis: If a LOOP should occur with these busses de-energized, many components needed to safely shutdown the plant would not be available. 0.0.2. Operating Procedures OP/0/A16350/OO1C, 250 VDC Auxiliary Power System contains fifteen enclosures which address the following activities:

Placing a battery charger in-service.
Removing a battery charger from service.
Performing an equalizing charge on a battery.
Removing a battery from service.

From Lesson Plan OP-MC-EL-EPJ Pg 17 (Rev 12) IMXF IMXH 2MXH 2MXF IDP IDS 2DS 2DP IDP 2DP BATTERY BATTERY TYPICAL TO LOADS TYPICAL TO LOADS I DPI2DP LOADS

1. TURBINE BIU VAPOR EXTRACTOR

2. GROUND AND UV DETECTORS

3. TURBINE EMER. BRG. OIL PUMP

4. GEN. AIR SIDE SEAL OIL BIU PUMP 5 FWPT A EMER OIL PUMP

6. FWPT B EMER. OIL PUMP

7. DLA (SERV. BLDG., CR, EQUIP. ROOM)

8. DLB (AUX. BLDG. DIG ROOM)

9. DLC (TURB. BLDG.)

10. DLD (RX BLDG.)

11. DLE (ADMIN. BLDG.)(IDP ONLY)

Parent Question MNS Bank ELEPJOO7 1 Pt Initial Conditions:

Both Units at 100% power.
All controls in AUTO.
All electrical busses aligned to their normal supplies except as follows:

o 250 VDC Auxiliary Supply Battery 1 DP is aligned for equalizing charge. Problem: The Standby Charger (1 DS) DC Output Breaker has tripped open due to a Charger internal fault. Which ONE of the following describes the current supply to 250 VDC Bus IDP? A. Chargers 1 DP and 2DP will be supplying the bus. B. Charger 1 DP will be the sole supply to the bus. C. Charger 1 DP and Battery 1 DP will be supplying the bus. D. Bus 1 DP will be de-energized. Answer A

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 23 SYSO64 A4.08 Emergency Diesel Generator (EDIG) System

ility to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) Jpening of the ring bus Given the following:

Uniti is shutdown in MODE 3
Auxiliary Transformer 1ATA is tagged out for repairs
All unit loads are being supplied by Auxiliary Transformer 1ATB

1. A Blackout will occur if open.

2. The DG Sequence is ONLY enabled if emergency bus minimum voltage and frequency setpoints are met.

Which ONE (1) of the following completes the statements above? A. 1. PCBs8&9

2. Committed B. 1. PCBsII&12

2. Committed C. 1. PCBs8&9

2. Accelerated D. 1. PCBs11&12

2. Accelerated Tuesday, July 13, 2010 Page 64 of 294

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2010 MNS SRO NRC Examination QUESTION 23 eneraI Discussion Since one busline is already out, loss of the other busline which feeds 1ATB will result in a Blackout on both 4160V busses. The busline which feeds 1ATB is fed from the switchyard via PCB 11 & 12. When bus voltage is greater than or equal to 92.5% and DIG speed is greater than or equal to 97%, the accelerated sequence is enabled. Blackout loads will be sequentially applied at intervals of approximately 2 seconds, as long as bus voltage remains greater than or equal to 92.5% and frequency remains>58.2Hz. Complete loading of all blackout loads, via the accelerated sequence, could be done in as little as 25 seconds. Should the Accelerated Sequence Relay scheme fail to work, the Committed Sequence would be actuated approximately 10 seconds after the diesel receives its blackout start signal if load shed of the bus has been completed. The committed sequence may take up to 12 minutes to load all blackout loads. The committed sequence does not require any minimum voltage or minimum frequency to allow it to progress as does the elerated Sequence. The Committed Sequence is required by Technical Specifications. Answer A Discussion above. PLAUSIBLE: Part I is plausible if the applicant does not recall which breakers feed which busline. [rt 2 is plausible if the applicant does not recall the difference between the .Accerated and Committed start sequences Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Part I is correct. [Part 2 is plausible if the applicant does not recall the difference between the .Accerated and Committed start sequences. Answer C Discussion (NCORRECT: See explanation above.

  ?LAUSIBLE: Part 1 is plausible if the applicant does not recall which breakers feed which busline.

Part 2 is correct. Answer D Discussion -- CORRECT. See explanation above. Basis for meeting the KA Opening of the ring bus equates to a loss of or malfunction in the switchyard. In this case the opening of the ring bus is the opening of Switchyard PCBs 11 & 12 which results in a Blackout on 1ETA and IETB. The ability to monitor portion of the KA related to opening of the ring bus and the Emergency Diesel Generator system is met by the applicant demonstrating a knowledge of how the EDGs operate under these conditions. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source L Development References Student References Provided Lesson Plan OP-MC-DG-EQB Pg 2 (Rev 16) C)PMCDGDG OBJ. #5 S064 A4.08 Emergency Diesel Generator (EDIG) System

Ability to manually operate and/or monitor in the control room: (CFR: 41.7/45.5 to 45.8) Opening of the ring bus Tuesday, July 13, 2010 Page 65 of 294

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2010 MNS SRO NRC Examination QUESTION 23 401-9 Comments: RemarkslStatus

401-9 Comments: An SI has not occurred. In order to give starting the EDO in priority mode some credibility, can you state in another bullet that SI has NOT been reset (because one has not occurred but the applicant will have to figure that out)? This will lend B2 and C2 additional credibility or put something in the stem that might indicate an SI might have occurred but didnt. This Q is E because of 2 weak distracters.. Resolution I Comments: Dont feel that theres a way to add the SI angle and maintain plausibility. Actually, two validators have picked B. Wrote a potential replacement question that asks the applicant to differentiate between the Commited and Accelerated Sequences. Tuesday, July 13, 2010 Page 66 of 294

Question 23

References:

NNLLL OBJECTIVE L L P P 0 OORSR ROO 1 State the purpose of the Diesel Generator Load Sequencing X X X X System. 2 List the Sequencer Automatic Actuation Signals. X X X X X 3 List the two Sequencer Modes of Operation and give a brief X X X X X explanation of each mode. 4 State which of the Sequencer Modes has priority. X X X X X 5 Describe the sequence of events which occur during the X X X Blackout Mode of Sequencer Operation. 6 Describe the sequence of events which occur during the X X X Safety Injection Mode of Sequencer Operation. 7 Describe the sequence of events which occur during a X X X Blackout followed by a Safety Injection. 8 Describe the sequence of events which occur during a Safety X X X Injection Actuation followed by a Blackout. (NOTE: with S reset and with S not reset). 9 Describe the sequence of events required to be done in order X X X to return the 4.16 KV bus back to normal following a:

       . Safety Injection
       . Blackout
       . Safety Injection followed by a Blackout
       . Blackout followed by a Safety Injection 10  Given a Limit and/or Precaution associated with an operating X  X  X  X  X procedure, discuss its bases and when the it applies.

From Lesson Plan OP-MC-DG-EQB Should a Safety Injection signal occur at any time after the first time delay relay completes its cycle, the circuit will automatically initiate separation from the offsite power source and transfer to the emergency diesel generators. Protection for a severe diesel-generator overload accompanied by a system voltage dip caused by events such as a Loss of Off-Site Power (LOOP) with the Diesel Generator operating in parallel with the grid is provided by the voltage-controlled overcurrent relay (51V). This relay consists of three single phase relays (51VX, 51VY, and 51VZ). The operation of any one of these phase relays will activate an annunciator alarm in the control room (AD-il A-4 (D-4), D/G A (B) Overcurrent) to warn the operator of an overload condition (800 amps @ 3360 volts). Operation of any two of these overcurrent relays will result in operation of the diesel-generator lockout relay (86D). Diesel-generator lockout relay (86D) will trip and lockout the diesel-generator switchgear breaker and initiate a shutdown of the diesel-generator. This lockout must be reset by hand before the breaker can be reclosed. 3.3 Sequencer operation during a Blackout I Objective # 5 Sequencer operation during a Blackout with no safety injection signal and the under-voltage is ! due to fault relay 86N, 86S or 86B. If 213 LOV Relays sense a loss of voltage on their associated 41 60V bus, the blackout relay will pick up and actuate a DIG start, If the UV condition still exists after 8.5 seconds, the blackout logic is sealed in. All 4160V breakers on the bus are then tripped open. When DIG speed is 95%, the output breaker will close. When bus voltage is 92.5% and DIG speed is 97%, the accelerated sequence is enabled. Blackout loads will be sequentially applied at intervals of approximately 2 seconds, as long as bus voltage remains 92.5% and frequency remains> 58.2 Hz. Complete loading of all blackout loads, via the accelerated sequence, could be done in as little as 25 seconds. If during the sequencing of blackout loads the Sequencer RESET pushbuttons are depressed, no additional sequencing will occur. This is because once the RESET pushbuttons are depressed, the blackout signal is removed and since there is power on the 4160V bus a blackout no longer exists. It would require another blackout signal or manual loading of the bus to complete the sequencing of loads. Should the Accelerated Sequence Relay scheme fail to work, the Committed Sequence would be actuated approximately 10 seconds after the diesel receives its blackout start signal if load shed of the bus has been completed. The committed sequence may take up to 12 minutes to load all blackout loads. The committed sequence does not require any minimum voltage or minimum frequency to allow it to progress as does the Accelerated Sequence. The Committed Sequence is required by Technical Specifications.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 24 SYSO73 Kl.O1 Process Radiat

                      -               ion Monitoring (PRM) System f     nowledge of the physical connections and/or cause-effect relationships betwee n the PRM system and the following systems: (CFR: 41.2 to 1.9/ 45.7 to 45.8)

Those systems served by PRMs Which ONE (1) of the following lists the EMFs that will automatically stop the Auxiliary Building Unfiftered Exhaust Fans (1ABFXF-1AI1 B) on a Trip 2 alarm? A. 1 EMF 36(L) Unit Vent Gas (Low Range) OR 1 EMF 36(H) Unit Vent Gas (High Range) B. I EMF 36(L) Unit Vent Gas (Low Range) OR 1EMF 37 Unit Vent Iodine C. 1 EMF 35(L) Unit Vent Particulate (Low Range) OR 1 EMF 37 Unit Vent Iodine D. 1 EMF 35(L) Unit Vent Particulate (Low Range) OR 1 EMF 36(H) Unit Vent Gas (High Range) Tuesday, July 13, 2010 Page 67 of 294

___ __ __________ __ _____ __________ __ _____ __________ _____ FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 24 EC General Discussion The automatic actions for the Unit Vent EMFs are as follows:

1) A Trip 2 high radiation alarm on 1EMF 35 (L), IEMF 37, 2EMF 35 (L), or 2EMF 37 will stop Auxiliary Building Unfiltered Exhaust Fans IABFXF-1A, IABFXF-IB, 2ABFXF-1A, and 2ABFXF-2B.

2) A Trip 2 high radiation alarm on 1EMF 36 (L) will close 1WG16O to terminate waste gas discharge.

   ,3)IEMF 36 (L) will also alarm and indicate at the Waste Gas Processing Panel.

Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because both EMFs monitor the Unit Vent and both have automatic actions (IEMF 36(L) closes IWG16O and 1EMF 36(H) stops the sample pump). Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because IEMF 37 is correct and IEMF 36(L) monitors the Unit Vent and has automatic actions (closes I

--

Answer C Discussion

                             --
                                        -_

__ I Sanon Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because 1EMF 35(L) is correct and 1EMF 36(H) monitors the Unit Vent and has automatic actions (stops the sample pump on a Trip 1 alarm). Basis for meeting the KA The KA is matched because the applicant must know the automatic actions (cause-effect relationship) that occur on a Trip 2 alarm for the EMFs ( monftoUnft Vent.__ 3asis for Hi Cog Basis for SRO only JobLeveTCognWveLeveltTyp___ estionSourcT Development References esson man OP-MC-WE-EMF Section 2.14 SYSO73 K1.01 Process Radiation Monitoring (PRM) System 41.9/45.7 to 45.8)

z Knowledge of the physical connections and/or cause-effect relationships between the PRM

system and the following systems: (CFR: 41.2 to Those systems served by PRMs Remarks!Status 401-9 Comments: It would appear from the reference that B is also a correct answer since since 1EMF 36(L) and 1 EMF37 are both called out as tripping mechanisms. The reference does NOT discern between particulate or gas. The Q is U until facility re-verifies because of 2 potentially correct answers. Tuesday, July 13, 2010 Page 68 of 294

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2010 MNS SRO NRC Examination QUESTION 24 lution/Coents: A Trip 2 alarm on 1EMF 36(L) will ONLY close 1WGI6O. It will NOT stop the Auxiliary Building Unfiltered Exhaust Fans. Likewise, a Trip 2 alarm on 1EMF 35(L) or 1 EMF 37 will [ONLY stop the Auxiliary Building Unfiltered Exhaust Fans. It will NOT close 1WG 160. They are two separate actions from [different EMFs. Therefore B cannot be correct because 1EMF 36(L) will not stop the Auxiliary Building Unfiltered Exhaust [Fans. Tuesday, July 13, 2010 Page 69 of 294

Question 24

References:

From Lesson Plan OP-MC-WE-EMF Section 2.1.4: The purpose of the auto actions:

EMF34 effluent is directed to ground water drainage sump A, therefore isolating this flowpath prevents contaminating this sump.
S/G blowdown blowoff tank effluent may be directed to either the condensate system or the turbine building sump, isolating blowdown will prevent contaminating these systems via the blowdown pathway.
Conventional sampling effluent may be directed to the CST or turbine building sump, isolating conventional sampling will prevent contaminating these systems via this pathway.

These channels use dual range gamma liquid assembly. The low range uses a gamma liquid ( Nal Scint) while the high range uses a GM detector. 2.1.4 Unit Vent Airborne Monitor The following channels are used to monitor the unit vent:

1(2) EMF 35 (L) Unit 1(2) Unit Vent Particulate (Low Range)
1(2) EMF 36 (L) Unit 1(2) Unit Vent Gas (Low Range)
1(2) EMF 36 (H) Unit 1(2) Unit Vent Gas (High Range)
1(2) EMF 37 Unit 1(2) Unit Vent Iodine Objective # 2 These EMFs, utilize a sample probe located within the Unit Vent to monitor, record, and alarm the gaseous, iodine and air particulate activity levels released to the atmosphere from the combined ventilation systems within the station.

Atmosphere from the Containment Purge, Containment Annulus Ventilation, Auxiliary Building Ventilation, Condenser Air Ejector, Fuel Pool Ventilation and other potentially radioactive systems are discharged through the Unit Vent. Objective # 2, 3 The automatic actions for these EMFs are as follows:

A Trip 2 high radiation alarm on 1EMF 35 (1), 1EMF 37, 2EMF 35 (1), or 2EMF 37 will stop Auxiliary Building Unfiltered Exhaust Fans 1ABFXF-1A, 1ABFXF-1B, 2ABFXF-1A, and 2ABFXF-2B

A Trip 2 high radiation alarm on 1EMF 36 (L) will close 1WG16O to terminate waste gas discharge.
1EMF 36 (1) will also alarm and indicate at the Waste Gas Processing Panel.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 25 SYSO76 K3.07 Service

                  -           Water System (SWS) nowledge of the effect that a loss or malfunction of the SWS will have on the following:

(CFR: 41.7 I 45.6) SF loads Which ONE (1) of the following is an effect if flow is lost to the Nuclear Servic e Water System Essential Header? (Assume the equipment listed is in service) A. PD pump bearing oil temperature increases. B. NC Pump motor bearing temperature increases. C. MD CA Pump motor bearing temperature increases. D. Steam Generator Blowdown Heat Exchanger outlet temperature increas es. Tuesday, July 13, 2010 Page 70 of 294

_____ ___ _____ _____ _____ _____ ____ _____ _____ ___ _____ _____ _____ ____ _____ _____ __ ____ _____ _____ FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 25 eneral Discussion The MD CA Pump motor cooler is one of the loads supplied by the RN Essent ial Header. Answer A Discussion i1CORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant confuses the PD pump bearing oil cooler with the NV pump bearing oil coolers which pplied from the RN Essential Header. are Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because the NC Pump motor coolers are supplied with RN except it is from the Non-Essential Heade is also plausible that the applicant may believe the NC Pump motor coolers to rs. It be an Essential Header load since on an SI the RN supply to all LNon-Essential loads is isolated with the exception of the NC Pump Motor coolers. Answer C Discussion 3iERRECT: See explanation above. Answer D Discussion JCORRECT: See explanation above. IBirswer is plausible because the SGBD HXs are supplie d by RN except they were are supplied by the Non-Essential Heade Basis for meeting the KA r.j iiis matched because the applicant must know the ESF loads that are supplied by Nuclear Service Water to correctly answer the questio n Basis for Hi Cog Basis_for SRO only Level Cognitive Level Question Source RO Memory BANK J MNS Exam Bank Question #PSSRNO21 Development References tu dent References Provided Lesson Plan OP-MC-PSS-RN Section 2.4 and 3.2.3 earning Objective OP-MC-PSS-RN #10 SYSO76 K3.07 - Service Water System (SWS) Knowledge of the effect that a loss or malfunction of the SWS will have on the following: (CFR: 41.7 / 45.6) ESF loads RemarkslStatus 401-9 Comments: No comment. Resolution / Comments: Tuesday, July 13, 2010 Page 71 of 294

Question 25

References:

From Lesson Plan OP-MC-PSS-RN Section 2.4: 2.4 Supply Headers Objective # 8, 10 The RN system provides flow to the following headers:

A and B essential headers
Reactor building Non-essential headers
Auxiliary Building Non- essential headers There is one redundant essential header for each train. These headers contain the equipment and component essential for safe shutdown of the plant. (Refer to Drawing 7.4 and 7.6). The following loads are supplied by the essential header. Included in this listing is whether these components are supplied on S, BC or S:

LOADS S BC *5

1) Pump motor coolers/AHU

Component Cooling Pump motor ( KC) X X
Centrifugal Charging Pump motor ( NV) X X
Safety Injection Pump motor ( NI) X
Residual Heat Removal Pump motor ( ND) X
Containment Spray Pump motor ( NS) X
Fuel Pool Cooling Pump motor ( KF) X X
Nuclear Service Water Pump motor ( RN) X X
Auxiliary Feedwater Pump ( CA) X X

2) Heat exchangers:

Containment Spray ( NS)

X

ComponentCooling(KC) X X
Diesel Generator Engine Cooling ( KD) X X
D/G Starting Air Compressor After Cooler X X
Control, Cable and Equip Room A/C Cond (YC) X X

3) Oil coolers

Centrifugal Charging Pump Bearing (NV) X X
Centrifugal Charging Pump Gear ( NV) X X
Safety Injection Pump Bearing ( NI) X
Note: If an S is present you will also have had an S. The S column only shows the additional loads which would be served when the Sp occurs.

4) Supplies assured makeup for the following systems:

Auxiliary Feedwater (CA)
Component Cooling (KC)
Spent Fuel Pool Cooling (KF)
Diesel Generator Cooling (KD)

The RN return from the NS heat exchangers is monitored for radioac tivity by EMF-45A & B to detect tube leakage. The NS heat exchangers have a wet lay-up loop associa ted with the sheilside ( RN ) of the heat exchanger ( Refer to Drawing 7.7). This wet lay-up loop was added to help reduce corrosion buildup on the sheliside of the HX. The 2B NS heat exchanger wet lay up loop is on the tube (RN) side of the heat exchanger. This system is non-safety related and in case of a break in the system there are flow limiting orifices on the suction and discharge sides. This system is primarily the responsibility of the Chemistry Dept. with the exception of the isolation valves directly off the RN piping which will be Operations. The wet lay up system will normally be in service with the isolation valves open and the heat exchanger water solid. The recirc pump will be run for sampli ng purposes and chemical additions as necessary. The RN Reactor Building non-essential header is not redundant and is isolated on an S ( Phase B) signal, when it is being supplied from the A RN header. If B train is supplying the header, flow will be lost to the NCP coolers on a BO or SS. This header contains the NCP motor coolers ( Refer to Drawing 7.6). Loss of RN to the NCP motor cooler(s) requires the operator to trip the effected NCP(s). Objective # 11 The RN Auxiliary Building non-essential header is not redundant and is isolated on an S signal. The components supplied by this header are: (refer to Drawing 7.6)

Reciprocating Charging Pump Bearing oil cooler
Reciprocating Charging Pump Fluid Drive oil cooler Note: The Steam Generator Blowdown Heat Exchanger has been flanged out and abandoned in place for Unit #1 ( NSM 12430) and Unit #2 ( NSM 22430).

Due to both units alignment to the RL Header, a cross-tie is created between the units through a 6 inch line. (Refer to drawing 7.4) The reason that the Auxiliary Building non-essential header supply isolatio n valve (RN42 ) is NOT closed during a Blackout is to allow A RN pump supply the Reactor Building ventilation units ( refer to Drawing 7.11). The A RN pump will have a greater NPSH since it will be supplied by the LLI. Also it is likely under Blackout conditions the RV pumps will not have power. Due to fouling problems and repeated maintenance on the PD pump heat exchanger a decision was made to isolate the Aux. Bldg. non-essential header. As a result the normal position of 1RN-64 will be closed. When it is necessary to start/stop the PD pump 1RN -64 will be opened/closed per the NV procedure.

From Lesson Plan OP-MC-PSS-RN Section 32.3: 3.2.3 Safety Injection Alignment On receipt of a Safety Injection signal basically the same autom atic actuation occurs as after a blackout. The exceptions are that the supply to all nonessential equipm ent except the NC pump motor coolers and crossovers between essential trains are isolated. The A RN pump supplies Reactor Building non T essential header. The RV pumps will start automatically and supply the containment ventilation units if a blackout does not occur concurrently with the LOCA. Drawings 7.14 and 7.15 provides the flow path for a unit safety injection. NOTE: An 5 S signal will affect both units suction, discharge and AB non-es sential headers. Refer to Drawing 7.14 On receipt of a Phase B isolation signal (S the RV pump suction

                                               )                       is isolated to conserve water. The containment isolation valves close to isolate the NC pump motor coolers. All nonessential supply is isolated providing double isolation at this time between all essenti al and nonessential equipment. The NS heat exchanger inlet isolation valve is opened from the contro l room when required. During all modes of operation, water is available for assured makeup. Drawin gs 7.16 provides the flow path following a unit safety injection with a phase B signal.

4.0 TECHNICAL SPECIFICATIONS Objective # 17 4.1 Tech Spec 3.7.7 Nuclear Service Water System (NSWS) 4.2 Tech Spec 3.7.8 Standby Nuclear Service Water Pond (SNSWP)

PARENT QUESTION: PSSRNO21 1 Pt Which of the following is a Nuclear Service Water System essential header load? A. Steam generator blowdown heat exchanger. B. Auxiliary feedwater pump motor cooler. C. Reactor coolant pump motor winding cooler. D. Positive displacement pump bearing oil cooler. Answer 115 B PSS-RN, section 2.4 Objective 10

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 26 252 SYSO78 A4.O1 Instrument Air System (lAS)

bility to manually operate and/or monitor in the control room: (CFR: 41.7 /45.5 to 45.8) ressure gauges Due to a leak on the VI system the following indications were observed:

1AD-12 Cl (VIA/S Lo Pressure) is LIT
OVIP-5090 (VINS Press) dropped to a lowest reading of 86 PS1G and is now 89 PSIG and increasing Which ONE (1) of the following describes automatic actions which have occurred as a result of the indicated pressure transient?

A. G and H VI Compressors auto-started ONLY. B. 1VI-820 (Vito VS Supply) auto-closed ONLY. C. IVI-820 auto-closed AND IVI-1812 (VI Dryer Bypass Vlv) has auto-opened. D. G and H VI Compressors auto-started AND 1VI-820 (VI to VS Supply) auto-closed. Tuesday, July 13, 2010 Page 72 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 26 2526 0 eneral Discussion At a decreasing VI pressure of 90 PSIG the following actions occur: IVI-820 (VI to VS Supply) Auto closes G and H Compressors (Diesel VI compressors) Auto Start If VI pressure continues to decrease to 85 PSIG, 1VT-1812 (VI Dryer Bypass) will OPEN. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because this action would have occurred but is not complete. Answer is incomplete and incorrect due to the ONLY designation. Answer B Discussion INCORRECT: See explanation above. - -______________ PLAUSIBLE: This answer is plausible because this action would have occurred but is not complete. Answer is not complete and is incorrect due to the ONLY designation. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because the applicant may conclude that lVI-1812 actuates with the other components at 90 PSIG. The first part is correct. Answer D Discussion CORRECT: See explanation above. Basis for meeting the KA KA is matched because the candidate, given information obtained from monitoring a trend of VI pressure indications located in the control room, what automatic actions have occurred associated with the Instrument Air system. asis for Hi Cog

 ,is is an analysis level question because the applicant must evaluate a given set of plant conditions, must recall a setpoint from memory, an then compare the plant conditions to the recalled memory to eliminate distracters and determine if a set of automatic actions should have occurred.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK M1JS (Bank 2227) Development References Student References Provided Lesson Plan OP-MC-SS-VI Objective 7 Section 1.2.10 page 67 and Objective 2 Section 1.3.1 page 89 ARP for 1AD-12 Cl (VI/VS Low pressure) SYSO78 A4.01 Instrument Air System (lAS)

Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) Pressure gauges 401-9 Comments: Remarks!Status 401-9 Comments: No comment. Resolution / Comments: N/A Tuesday, July 13, 2010 Page 73 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 26 Tuesday, July 13, 2010 Page 74 of 294

Question 26

References:

NNLLL OBJECTIVE L L P P 0 OORSR ROO 6 Explain the control function associated with each of the X X X X following VI Air Compressor (A, B, and C) pushbuttons:

     . Start/Stop pushbutton
     . Reset pushbutton List the interlocks / trips associated with operation of the   x x x x x following plant air system components:
   . VI Air Compressors
   . VI-820 (VI to VS Supply Valve)
   . VS Low Pressure Air Compressor
   . VBAir Compressor 8  Describe the following controls and/or indications associated   X X X X with operation of VI Air Compressors D, E, and F:
   . On/Off switch and indication
   . Start/Stop pushbuttons
   . Pre-lube pump status
   . Acknowledge/Reset pushbutton 9  Describe how the following VI System components function to     X X X X provide a continuous supply of clean dry air:
   . Service Building Air Receiver Tanks (and drains)
   . Air Dryers
   . Auxiliary Building Instrument Air Tanks 10  Explain each one of the following controls and /or indications, X X X X associated with the Breathing Air Compressors:

Start/Stop Pushbutton

   . Power ON Light
  . RUN Light
  . Discharge Air Over-Temperature Light
  . Rotor Oil Filter Service Light
  . Bearing Oil Filter Service Light
  . Air/Oil Separator Service Light
  . Service Air Filter AP Gauge
  . Purification Filter AP Gauge
  . Rotor Coolant Temperature Gauge
  . Discharge Air Pressure Gauge
  . Discharge Air Temperature Gauge 11 Describe normal operation of the Breathing Air                   X X X X X

Compressor(s). I From Lesson Plan OP-MC-SS-Vl Pg 71 (Rev 33) The VI System normally supplies the Low Pressure VS System through control valve 1 VI-820. Controls and indication for 1 VI-820 are located at the VI Sequencer Control Panel. The valve control switch is a three position switch:

Close
Auto
Open Objective # 7 Indication provided at the VI Sequencer Control Panel consists of the following:
1VI-820 Close (green light)
1VI-820 Open (red light)

This valve is normally in the AUTO position and will automatically close should VI System Pressure decrease to <90 psig. Upon valve closure IVI-820 can be reopened once VI System Pressure has increased >90 psig by placing the valve to the OPEN position. After opening the valve IVI-820, the switch should be returned to the AUTO position. If not, the valve will reopen without operator action, after closure, as soon as pressure has increased above 90 psig. 1.2.13 VI System Air Dryers A, B, and C Objective # 9 VI Dryers A, B, and C (AMLOC-CHA Dryers) are fully automatic, desiccant-type air dryers designed to remove vaporous moisture from the Instrument Air System. Generally, two of the three desiccant air dryers (A, B, and C) are in-service while one remains in standby, ready and available for service when needed. Each in service dryer will alternately cycle air through one of the two desiccant chambers for moisture removal, while the other chamber is regenerated (removal of previously adsorbed moisture) and re-pressurized.

From Lesson Plan OP-MC-SS-Vl Pg 75 (Rev 33) Purge Dump Restrictor Closes during dump and limits gas flow to prevent fluidization by controlling the rate of depressurization. Opens fully during all other periods. Dryer System Bypass Valve lVl-1812 is installed between the Dryer System Manual Bypass Valves 1VI-093 and lVl-094. This valve is designed to fail open on a loss of power or loss of air. Valves lVl-093 and lVl-094 will be normally open while 1Vl1812 will be normally closed. A solenoid operator associated with valve lVl-1812 is connected to pressure switch 0VlPS5381. The solenoid is set to vent the actuator upon receipt of a VI System Low Pressure epressure switch 07rPS5381 is cohnected to the]nstrume irrently controls the dryer Purge Exhaust Isolation V

  !AIJ L!1 UJIL 1VI-1 840) which fail closed on a low pressure signa 1];J.4;3i1             d to the REFLASPaneLsuch that.an ala[rjilr#               0
  ?1!JJ ii 11 Lfl irir E] yerPaneLIroub There is local indication of valve position, RESET and OVERRIDE capabilities provided at the Reflash Panel. By depressing RESET, 1VI-1812 will close, and by depressing OVERRIDE, 1VI-1 812 can be manually opened.

1VI-1812 is designed to automatically open and bypass the VI Dryers in the event of sudden blockage of flow due to some dryer malfunction. The PRA Study identified VI Dryer malfunctions as a primary contributor to Loss of VI event probability. The manual bypass valves (1VI-093 and 1VI-094) cannot protect against sudden dryer flow blockage events (e. g. switching valve failure). A filter is installed at the inlet of 1VI-1812 to prevent the potential of substantial contamination of the normally dry VI headers with rust known to exist in the wet VI headers. Instruments and Their Basic Function The A, B, and C VI Dryers are equipped with a set of gauges to indicate inlet air pressure, outlet air pressure, purge flow, and chamber pressure. The gauges are provided to monitor system operation. The gauges on the chamber indicate which chamber is on-stream (the gauge on the off-stream chamber should indicate zero (0) PSIG). The gauges are also used to verify that the internal pressure has been completely vented to the atmosphere when servicing is required. All pressure gauges should indicate zero (0) PSIG before any service work is performed on the dryer. Additional instruments include:

Chamber pressure relief valves.

Provide chamber protection if high pressure should develop during dryer operation. Set to relieve at design pressure.

Chamber Pressure Sensors Set to sense the lack or presence of chamber pressure following repressurization or depressurization.

From Lesson Plan OP-MC-SS-Vl Pg 67 (Rev 33) Objective # 4 The Diesel VI Compressors operate in two modes of operation. These modes are Automatic and Manual. In the Manual Mode of operation, an operator will start and run the compressor using controls on the compressor control panel located at the compressors themselves. For a manual start of the compressor to be accomplished, the following must be true:

The AUTO/OFF-RESET switch must be selected to the OFF-RESET position
The START/WARM-UP/RUN switch is in the WARM-UP Position
The HIGH/LOW switch is selected to the desired position (normally HIGH)

The operator then rotates the Engine Switch from the OFF position to the ON position and the compressor should start. Once the compressor has started and has warmed up, the operator can select the RUN position on the START/WARM UP/RUN selector switch to allow the compressor to load. If the operator is starting the compressor as directed from the Loss of Instrument Air System Abnormal Procedure, the AP directs the operator to leave the START/WARM UPIRUN switch in the RUN position to allow for immediate loading. The following is a set of conditions, which will allow the Diesel VI Compressors to automatically start:

The AUTO/OFF-RESET switch must be selected to AUTO
The START/WARM-UP/RUN switch is selected to RUN
The HIGH/LOW switch is selected to HIGH
The Latching Relay picks up The compressor will automatically start and load to the desired pressure.

Objective # 7 There are three signals, which will send an AUTO START signal to the Diesel Powered VI Compressors. These signals are:

Loss of VI header pressure as measured by OVIPS5O7O

   ** set at 90 psig decreasing
   + Compressor control can be regained when pressure increases above 95 psig

Loss of 313 KR flow to the D, E, and F VI Compressors
Loss of power to the VI Sequencer Panel (SKU#43) ISLXD/2SLXD-SMXU

MNS Bank Question 2227: Due to a leak on the VI system the Unit 1 OATC observes the following indications:

1AD-12 Cl (VINS Lo Pressure) is LIT
OVIP-5090 (VIIVS Press) dropped to a lowest reading of 86 PS1G and is now 89 PSIG and increasing Which ONE (1) of the following describes automatic actions which have occurred as a result of the indicated pressure transient?

A. G and H VI Compressors Auto Started ONLY B. lVl-820 (Vito VS Supply) Auto Closed ONLY C. 1VI-820 Auto Closed AND lVl-1812 (Vi Dryer Bypass Vlv) has Auto Opened D. G and H VI Compressors Auto Started AND 1VI-820 (Vito VS Supply) Auto Closed Th ANSWER: 0

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 27 SYSO78 K3.02 Instrument Air System (lAS)

(nowledge of the effect that a loss or malfunction of the lAS will have on the following: (CFR: 41.7 /45.6) Systems having pneumatic valves and controls Given the following:

Unft I is operating at 100% RTP when a loss of VI event occurs
AP-22 (Loss of VI) has been implemented
VI header pressure is 55 PSIG and decreasing Which ONE (1) of the following system effects would be the FIRST to require the crew to trip the reactor in accordance with AP-22?

A. Decreasing S/G levels B. Loss of RN supply to Containment C. Loss of NC pump seal leakoff to the VCT D. PZR level approaching the High Level Trip setpoint Tuesday, July 13, 2010 Page 75 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 27 L 2527

 ,éneral Discussion The CF control valves use 0 60# valve operating air. Depending on the nature of the problem with VI and considering line losses, etc., these
                              -

valves could start failing at 70# or more VI pressure as indicated in the control room. The operating philosophy regarding loss of Main Feedwater at power is to trip the reactor. This will prevent challenging the Lo-Lo S/G automatic reactor trip and will result in better initial conditions at the time of the manual trip. If the CF valves were to get to less than 25% open (for 30 sec or more) on 3 out of 4 S/Gs, an AMSAC could also be_generated. For most scenarios, its likely the operator will have manually tripped the reactor prior to this occurring. Answer A Discussion CORRECT: See explanation above. Answer B Discussion -________________________________________________ INCORRECT: See explanation above. PLAUSIBLE: 1RN-252B does fail closed which would result in a loss of NSW cooling to the U-l NCPs. This is a significant operational concern and left in this condition would result in the need to trip the reactor and secure the NCPs. It is therefore plausible but incorrect because this condition would not be an immediate threat. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: 1NV-34A fails open but if the applicant believes that the failure mode of this valve is closed this would result in a loss of D/P across the A NCP #1 seal and require an immediate reactor trip and pump shutdown. Answer 0 Discussion INCORRECT: See explanation above. PLAUSIBLE: 1NV-238 does fail open which would result in maximum charging flow. This would represent a longer term operational concern but would eventually result is challenging the PZR high level trip setpoint and is therefore plausible. Basis for meeting the KA K/A is address because the applicant must understand the effect of a Loss of VI will have on 4 different pneumatic valves and how this loss uld affect the systems containing these components. ______________ dasis for Hi Cog This is a hi cog question because it involves a level of analysis of given situation, apply system knowledge and predict an outcome. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK MNS Exam Bank Question AP22NO1 Development References Student References Provided AP-22 (Rev 28) Enc 12 AP-22 (Rev 28) Page 8 AP-22 Bacdground Document Page 15 OP-MC-AP-22 Obj. 5 SYSO78 K3.02 Instrument Air System (lAS)

Knowledge of the effect that a loss or malfunction of the lAS will have on the following: (CFR: 41.7 / 45.6) Systems having pneumatic valves and controls 401-9 Comments: RemarkslStatus 40 1-9 Comments: Distractor D is a long shot and I do not believe it is plausible especially since other, more pronounced reactor trip criteria exists. Consider replacing distractor D. D is NP. Tuesday, July 13, 2010 Page 76 of 294

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2010 MNS SRO NRC Examination QUESTION 27 2527 Resolution) Comments: Distracter D is not the strongest distracter. However, it is possible and is therefore plausible._Would like to keep this one. Tuesday, July 13, 2010 Page 77 of 294

Question 27

References:

OP-MC-AP-22 Obj. 5 OBJECTIVES NNLLL OBJECTIVE L L P P 0 OORSR R00 Concerning AP/1/A15500/022 (Loss of VI): X X X

       . State the purpose of the AP
       . Recognize the symptoms that would require implementation of the AP AP22001 2   Describe the mitigating strategies (major actions) contained in        X X X the procedure.

AP22002 3 Given scenarios describing accident events and plant X X X conditions, evaluate the basis for any caution, note, or step. AP22003 4 Given scenarios describing accident events and plant X X X conditions, evaluate conditions which require application of continuous action steps. AP22004 5 State the failure modes of the components listed in AP/22, X X X Enclosure 12 (Valve Failure Mode on Loss of Air).

From AP-22 Enclosure 12 pages 1 of 6 MNS LOSS OF VI PAGE NO. M -D. ifl Enclosure 12-Page 1 of 6 1NFT 1 R I ValveFailureModeonLossofAir NOTE The aes Isted in this enclosure fah at vais ai: pressures. 55 valves:

a. The fohowing 56 valves fail closed:

155-15(14 St Slowdown Corn Outside sd Controli 155-25:is St Slowdown Cont Outside Isol Control 1BS-38 (1C 5/0 Slowdown Cant Outside Isol Contro: 165-46 ilD 5:0 Slowdown Cont Outside sal Canto:: 166-54 (A 5/0 66 Cont Inside lso: lB6-6A6 5t 65 Cont Inside Iso:: 156-74 :C SO 56 Cont Inside Isol: 165-64 (D 5/0 65 Cont inside Isol: 166-123 14 5G Slowdown Throttle Control 166-124 lB 5;G Slowdown Throttle Control: 165-125 (iC 5/0 Slowdown Throttle Cont-oh 166-126(10 5/0 Slowdown Throttle Cont!oh.

2. CA valves:

a. The following CA a?es fai; opel:.

104-604 (IA CA Pump Dsch To 1A 5:0 Control: 1C-5bA 14 CA Pump DischTo 15 5:0 Control: 104-445 16 CA Pump Dsch To IC 5/0 Controli 104-40615 CAPurrp DischTo 1DS/G Controli 104-6446 :1 TO CA Pump Disch To 14 50 Control: 104-5245 (Ui TO CA Pump D:sci To 15 50 Control: 104-4846 (Ui TO CA Pump Disch To IC 5/0 Cont!o: 104-3646 (Ui TO CA Pump Cisci To 10 St Control:.

3. CF valves:

a. The foLowin; CF .alves fa I closes:

1CF-32A6 (14 St CF Contro: 1CF-23A6 :16 St CF Control: 1CF-2045 MC St CF Control 1CF-17A6 1,10 SO CF Control 1CF-10445 :14 5/0 CF Control Bypass: 1CF-13SA6 (15 St CF Control Bypass: 1CF-1OSAS MC SIC CF Control Bypass: 1CF-1DYAS :1D SIC CF Control 6ypass.

From AP-22 Enclosure 12 pages 3 of 6 MNS LOSS OF VI PAGE NO. Ap?1As5ac.:n 121 Enclosue 12 PageS of 6

Re UNIT 1 Valve Failure Mode on Loss of Air

8. NV valves:

a. The folc*wing NV daQes faL open:

            . 1NV-i6A (NV Suppi) To D NC Loop Isol)

            . 1NV-136 NV Supply To 4 NC Loop iso!:

            . 1 NV-34A A NC Pump Seal Return lsol:

            . iN/-SOB is NC Pump Seal Return Itoh

           . 1N/-664 C NC Pump Seai Return Isol:

           . 1NV-625 ID NC Pump Seal Return Isol)

           . 1NV-124 (Letdown Pressure Control:

           . 1N/-238 (Charging Line Flow Conol)

INV-241 (UI Seal Water lnj Flow Control:

           . iN/-237A :soric Acid To Blender Control:.

b. The fotoi.ing NV valves fail to the VCT pot Won:

1NV-276 Ecess LID Hx OtIt 2Vvay Cntrl:
1N:-I27A cUD H< Outlet 3Way Temp Cntrli
1NV-157A (NC Filters Ott 3-Way Cntrl

c. The folowing NV valves fail closed:

           . 1NV-IA INC LID isol To Reqen Hx:

           . 1NV-24 (NC LID Isol To Regen Hx:

           . 1NV-21. INV Spray To PZR Isol:

           . 1NV-246 (C NC Loop To Ext LID Hx lso:

           . 1NV-256 iC NC Loop To E:sLD Hx Ito!:

1NV-25B Ui Excess LD Hx Outlet Cntrl)

           . 1 NV-354 (variable LID Orihce Outlet Cont Isol:

           . 1NV-394 (.4 NC Pump Stancipipe Fill:

           . 1NV-SSB (B NC Pump Standpipe Fill:

           . INV-714 (C NC Pump Standpipe Fill)

           . 1NV-575 (D NC Pump Standpipe Fill,:

           . 1N/-924 (NC Pumps Seal Byp Return Hdr Isol)

           . 1NV-121 (Ui ND Letdown Control:

1NV-i6Th (VCT .ent To WG 1501

       . 1NV-171A(BABlenderToCT Inleti

           . 1 NV-I 75.4 iSA Blender to i CT Outlet)

           . 1NV-457A (45 GPM LID Orifice Cutlet Contisol:

1N7458A :75 GPM LID Orifice Cutlet Contlsol)

I . 1 N/459 Ui Variable LID Orifice Outlet Flow Cntrl I

1NV-E.404 (IJ 1 ND To P:r Aux Spray Controli.

From AP-22 Enclosure 12 pages 3 of 6 MNS LOSS OF VI PAGE NO. Enclosiwe 12-Page$of6 UNIT I I Valve Failure Mode on Loss of AIr

9. RE valves:

a. The fol:owing RE valve fails closed:

1RF-821A Unit 1 RE Cont Outside Isol:.

13. RN valves:

a. The fobwin; RN valves fail open:

1RN-&9A {RN 10 A KC ft Conn: 1RN-1DSA A NV Pump Cooer Sup sol 1RN-114A A NI Pvm2 Cooler Sup soIl 1RN-126A (A NS Pump ESS AhJ Sup lso:1 1RN-1SDA A ND Pump ESS AHU Sup Isol) 1RN-140A A RE Pump ESS AHU Sup Isol) 1RN-190B RN To B KG Hx Conrol: 1RN-204B i:B N. Pump Cooler Sup lsoh 1RN-215B (B NI Pump Cooler Sup IscIl 1RN-27B B NS Pump ESS AHU Sup lsou 1RN-2316 B ND Pump ESS AHU Sup lsol:i 1RN-240B (B RE Pump ESS AHU Sup lsoh.

h. The folowhng RN valves fail case::

1RN-2iA 1A RN Straner Eackwasn Automatic Supply isoli 1RN-22A 1A RN Strainer Backwash Auton-,am: Drain 1RN-25B itS RN Strainer Backwash Automatic Supply isol*i 1RN-26B 1B RN Strainer Backwash Automatic Drain 1RN-2SDB Rb Non Ess Sup Cart Outside Isol, 1RN-277B (RB Non Ess Ret Cant Outside lsol*

11. Rvvalves:

a. The following Ri waives fail closed:

1R.-79A UI U AHUS Ri Cant Outsice Supply Hdr sal:
1Rf-lUiA. :Ui lU AHUS R. Cant Inside Return Hdrlscl)

            . 1R.;-SOB (Ut VU AHUS Ri Corn Inside Supply I-Icr Isoli

1Rl-1Q2B(U1 \U AHUS Ri Cant Outs:de Return Hdr Isol:.

From AP-22 Page 8 of 121: MNS LOSS OF /l p:3E NO. S of 121 UNIT 1 AT:o7JEx;E:E: P.ZSONSE Z5CNE NOT DET.:N:

11. Ccntinueo:

rn. Control NO temperature as fcIows: Throttle ND flow. NOTE

KO to ND Hx flow should be close to flow prio: to loss of VI. since it is normally controlleo by motor operated aIves.

a KO to ND HX flow indications fails low during a loss of VI. iteria1e indications are awailable at the following locations, if needed: 1A: 1KCFT-57Oiaw: bldg. 733 +2, .esi of cc.Linin MI-E4: l: 1KCFT-5680 au; tllcçi. 733 4. west s:e of column JJ-55L

               . IF NC temperature is greater than 200F, THEN maintaifl KC fow to ND Hx greater than 2000 GPM.

a Throte KC Flow to ND Hx as reqwred.

12. IF AT ANY TIME VI pressure is less than 70 PSIG. THEN align B Train RN to SNSWP PER Enclosure 7 (Aligning B Train RN to Pond).

NOTE CF Control \al,es will fail closed on ow VI presue. whch y result n Ar5AC actuatio, and Lo Lo 5/0 leeL

13. Check SIG levels AT PROGRAMMED

                               -                                    IF S/G leveLs are going down in an LEVEL                                                     uncontrolled manner, TIit.4 perform the following:

a. Trip reactor.

b. Contnue with this procedure as time

c. GO TO EPl/A&JDD:E-O i;Reactor Trip or Safety lnjeconl.

AP-22 Background document Page 15 STEP 13: PURPOSE: Prompt the operators to watch S/G levels because the CF control valves fail closed on a loss of VI. If SIG levels cant be controlled, the Operator is directed to trip the reactor. DISCUSSION: The CF control valves use 0 60# valve operating air. Depending on the nature of the

problem with VI and considering line losses, etc., these valves could start failing at 70# or more VI pressure as indicated in the control room. The operating philosophy regarding loss of Main Feedwater at power is to trip the reactor. This will prevent challenging the Lo-Lo SIG automatic reactor trip and will result in better initial conditions at the time of the manual trip. Refer to PIP 2-M-87-0208 where a automatic reactor trip occurred 5 mm after loss of offsite power due to loss of VI to the CF valves. If the CF valves were to get to less than 25% open (for 30 sec or more) on 3 out of 4 SIGs, an AMSAC could also be generated. For most scenarios, its likely the operator will have manually tripped the reactor prior to this occurring.

REFERENCES:

PIP 2-M-87-0208

Parent Question AP22NOI: Question 6 AP22N01 1 Pt Unit 1 is operating at 100 % power when a loss of VI event occurs. AP/1/A/5500/22 (Loss of VI) has been implemented. VI header pressure is 55 psig and going down. Which of the following conditions would initially jeopardize the plant and require the SRO to direct tripping the Unit 1 Reactor per AP111A15500122 (Loss of VI)? A. 1NV-238 (Charging Line Flow Control) fails closed. B. 1CF-23AB (B SIG CF Control Vlv) fails closed. C. 1RN-252B (RB Non Ess Sup Cont Outside Isol) fails closed. D. 1 NV-34A (A NC Pump Return Isolation) fails closed. Answer 6 B

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 28 25281 SYS 103 A4.04 Containment System

                  -
\bility to manually operate and/or monitor in the control room: (CFR: 4L7 / 45.5 to 45.8) hase A and phase 13 resets Given the following conditions on Unit 1:

A LOCA has occurred inside Containment
Containment pressure is currently 3.5 PSIG Which ONE (1) of the following describes the MINIMUM steps required before KC can be restored to Containment?

A. Reset Phase A B. Reset Phase B C. Reduce Containment pressure below 1.0 PSIG, reset Phase A D. Reduce Containment pressure below 3.0 PSIG, reset Phase B Tuesday, July 13, 2010 Page 78 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 28 2528 Jeneral Discussion Phase B actuation secures Component Cooling Water (KC) to the Reactor Coolant pumps, Nuclear Service Water (RN) to the Reactor Coolant Pump Motor Coolers, Containment Ventilation Cooling Water (RV) and Instrument Air (VI) to the containment. ase B can be reset with signal still present, once resets are pushed, we regain control of valves that close on the Phase B signal. Answer A Discussion --____________________________ INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant does not recall which signal (Phase A or Phase B) closes the Containment KC valves. Answer B Discussion CORRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: The answer if the applicant does not recall whether the KC valves are closed by a Phase A or Phase B signal. If the applicant concludes that the valves are closed by a Phase A signal it is reasonable to also conclude that Containment pressure must be reduced to less than 1.0 PSIG (where a Phase A signal would be initiated by the Hi Containment pressure SI) in order to reset the the Phase A signal. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible since the KC valves are closed by a Phase B signal and the signal must be reset to open the valves. It is reasonable for the applicant to conclude that the Hi-Hi Containment pressure signal seals in which would prevent resetting the Phase B signal unless Containment pressure is reduced to less than 3.0 PSIG. Basis for meeting the KA (nstrating a knowledge of when the Phase B reset must be operated to regain control of equipment operated by the Phase B signal, the pplicant demonstrates the ability to operate Phase B resets from the Control Room. Therefore the KA is matched. asis for Hi Cog Basis for SRO only Job Level RO Cognitive Level Memory T QuestionType BANK Question Source MNS Exam Bank ECCISENO4 Development References dent References Provided Learning Objective:

1) ECC-ISE#13

References:

1) Lesson Plan OP-MC-ECC-ISE Section 3.1 SYS 103 A4.04 Containment System

Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) Phase A and phase B resets 401-9 Comments: RemarkslStatus 40 1-9 Comments: Since Phase B is actuated (3.5 psig), it would appear that the KC valves are closed as stated. The distractor analysis for D appears to indicate that Cntmt pressure must be reduced to < 3 psig before phase B can be reset (because phase A is still in). In other words can one reset phase B without phase A being reset? Please re-verify this because the reference was not clear on this issue. Tuesday, July 13, 2010 Page 79 of 294

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2010 MNS SRO NRC Examination QUESTION 28 2528 This Q is E until re-verified. Resolution / Comments: Phase A does not have to he reset in order to reset Phase B. And, Phase B can be reset with containment pressure greater than 3.0 psig. The wording in distracter D analysis is incorrect. Instead of prevent resetting the Phase A signal unless Containment pressure is reduced is should have said prevent resetting the Phase B signal unless Containment pressure is reduced. [çged analysis for distracter D accordingly. Tuesday, July 13, 2010 Page 80 of 294

Question 28

References:

From Lesson Plan OP-MC-ECC-ISE Section 3.1: I Objective#13 I Phase B Containment Isolation is actuated by: 2/4 Hi Hi Containment Pressure > 3.0 psig on channels Manually 1/2 pushbuffons Phase B actuation secures Component Cooling Water (KC) to the Reactor Coolant pumps, Nuclear Service Water (RN) to the Reactor Coolant Pump Motor Coolers, Containment Ventilation Cooling Water (RV) and Instrument Air (VI) to the containment. Phase B can be reset with signal still present, once resets are pushed, we regain control of valves that close on the Phase B signal. Containment Ventilation Isolation (Sf4 is initiated by any of the following:

Safety Injection (Se)
Manual Phase A (St)
Manual NSIPhase B
Trip 2 alarm on EMF-38, 39, or 40 Containment Ventilation Isolation (SH) signal secures VQ and VP.

To Reset Containment Ventilation Isolation following a Safety Injection, Manual Phase A, or Manual Phase B, the Containment Ventilation (SH) Reset Pushbuttons must be depressed (can reset without resetting the initiating signal). To Reset Containment Ventilation following an EMF 38, 39, 40 Trip II, the EMF must be reset, then the Containment Ventilation Reset Pushbuttons must be depressed. NOTE: Resetting the SH signal will allow manual control of VQ valves. VQ valves do not have an auto function.

Annulus Ventilation System (VE) start maintains negative pressure in annulus. It is actuated automatically by a Hi Hi Containment pressure signal or manually by either depressing Manual NSIPhase B Pushbutton or placing VE (Annulus Ventilation) to ON. To reset the start signal we must reset the Phase B isolation, then, place VE (Ann ulus Ventilation) fan switch to Reset and place back in auto. [-12 Skimmer and Air Return Fan (VX) starts on a Hi Hi Containment Pressure (Sr) with CPCS or Manually by NS/Phase B pushbutton and CPCS after a 10 minute time delay.

Question 28 Parent Question: ECCISENO4 1 Pt Given the following conditions:

1) Containment pressure is 3.8 psig

2) Phase B containment isolation has occurred What are the minimum steps required to restore Component Cooling water to containment?

A. Restore KC to operation immediately B. Reset Phase B, restore KC to operation C. Reset SI, reset Phase B, restore KC to operation D. Reduce containment pressure below 3.5 psig, reset Phase B, restore KC to operation Answer 599 Answer B MISCINFO: RO&SRO SOURCE: BCH

REFERENCES:

OP-MC-ECC-ISE page 29 LESSON: OP-MC-ECC-ISE TASK: OBJECTIVE: 1.N.2 TIME: K/A: 022000K403 (3.6*/4.0*) DATE: 11/29/95

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 29 2529 SYSOOI K6.13 - Control Rod Drive System nowledge of the effect of a loss or malfunction on the following CRDS components: (CFR: 4 1.7/45.7) ocation and operation of RPIS Give the following conditions on Unit 1:

The unit is in MODE 3 withdrawing SID banks in preparation for startup
1AD-2 / D1O (RPI Urgent Alarm) Annunciator has just alarmed
DRPI and OAC RODS position indication for rod D-8 has been lost What is the FIRST action required by SLC 16.7.9 (Rod Position Indication System -

Shutdown)? A. Place rods in manual ONLY. B. Place rods in manual AND drive all rods in. C. Immediately open the reactor trip breakers. D. Restore rod position indication within 1 hour. Tuesday, July 13, 2010 Page 81 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 29 2529

  .,eneral Discussion SLC 16.9.7 (Rod Position Indication System Shutdown) requires that at least one rod position indicator be operable and capable of
                                                  -

determining the control rod position within + 12 steps for each rod not fully inserted. This SLC is applicable to Modes 3,4,5. In the situation given in this question, the unit is in Mode 3 in the process of withdrawing S/D Banks. If rod position is lost for any rod, Condition A requires that the Reactor Trip breakers be opened immediately. - Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: With any malfunction involving the control rods this would be the required action in AP-14. It In this case given this is not the orrect action because it is not required by SLC 16.9.7. -______________________

-;

Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: If the applicant correctly remembers that a shutdown is required but confuses the required action with a one hour requirement. The verification of shutdown margin is consistent with almost every 1 hour action statement concerning rod alignment and position indication with the unit in Mode 1 or 2 therefore it would be plausible for the applicant to apply that requirement to this situation. Answer C Discussion CORRECT: See explanation above. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: This would satisfy the requirements of TS 3.1.4 (Rod Group alignment limit Action B. With One rod not within alignment limits, Action B. 1 requires the rod to be restored within alignment limits within 1 hour. The applicant may incorrectly apply the actions of this spec because with the rod position indication unavailable it would be impossible to prove that it was within alignment limits. Basis for meeting the KA dthough there is no physical cause/effect relationship between the RPIS and CRDS, for this particular instance, a malfunction has occurred in ( e RPIS and the effect on the CRDS is that operator action is required by SLC 16.7.9 to immediately de-energize the CRDS. Therefore, the KA s matched. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK Bank M1JS ICEDAROI Development References ent References Provided - SLC16.9.7 SYSOO1 K6.13 Control Rod Drive System

L Knowledge of the effect of a loss or malfunction on the following CRDS components: (CFR: 4 1.7/45.7) Location and operation of RPIS 401-9 Comments: Remarks!Status 40 1-9 Comments: B is NP as written. Place in the stem what is the FIRST action and remove and do not move them from distractor B. E because distractor B is NP as written. Resolution / Comments: Revised question per Lead Examiners recommendation. Then Tuesday, July 13, 2010 Page 82 of 294

__ ___ FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 29 2529 rearranged distracters A and B for psychometrics mading B the new correct answer. If this is acceptable the distracter analysis will need to be reworked. See attached file for proped sion. Tuesday, July 13, 2010 Page 83 of 294

Question 29

References:

OP-MC-IC-EDA Obj. 10

10. Concerning the Technical Specifications related to the DRPI System:

Given the LCO title, state the LCO (including any COLR values) and applicability.
For any LCOs that have action required within one hour, state x x x the action.
Given a set of parameter values or system conditions, determine if any Tech. Spec. LCOs is (are) not met and any action(s) required within one hour. X
Discus the bases for a given Tech. Spec. LCO or Safety Limit

                               *SROONLY x x x x

From Selected Licensee Commitment 16.9.7 16.7 INSTRUMENTATION 16.7.9 Rod Position Indication System Shutdown

COMMITMENT One rod position indicator (excluding demand position indication) shall be OPERABLE and capable of determining the control rod position within +/- 12 steps for each shutdown or control rod not fully inserted. APPLICABILITY MODES 3, 4 and 5 with the reactor trip breakers in the closed position with rods not fully inserted and capable of withdrawal. NOTE For testing or trouble shooting, alternate methods may be used to ensure there is no possibility of rod motion. These methods are pulling fuses, sliding links in the rod control cabinets or removal of CRDM head cables. After one of these alternate methods is used, the reactor trip breakers may remain in the closed position. REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required Open the reactor trip breakers. Immediately rod position indicators inoperable.

Parent Question ICEDAROI Question 29 ICEDARO1 1 Pt(s) Unit #1 is in Mode 3. While withdrawing Shutdown Bank E, the DRPI rod position indication for Rod D-8 was lost at 96 steps. The Rod Position Indication (RPI) urgent failure annunciator, General Warning for D-8, and Rod Bottom Light for D-8 were received. OAC Program General 76 does not update for Rod D-8 when Bank E is moved. Select the action which must be taken by the operator: A. Immediately trip the reactor B. Place rods in manual and do not move them C. Continue the startup but do not enter Mode 1 D. Drive all rods in and verify shutdown margin within 1 hour Answer 29 A

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 30 253O SYSOI1 K302 Pressurizer Level Control System (PZR LCS)

nowledge of the effect that a loss or malfunction of the PZR LCS will have on the following: (CFR: 41.7 / 45.6) {CS Given the following conditions on Unit 1:

The unit is at 100% RTP
All Pressurizer heaters are energized in MANUAL
The SLIM for 1 NV-238 (Charging Flow Control) has been placed in MANUAL due to a malfunction of the Pressurizer Level Master Controller
The OATC reduces the 1NV-238 SLIM output to reduce Pressurizer level
Charging Line Flow is inadvertently reduced to 18 GPM If the INV-238 controller output remains constant, after 5 minutes Pressurizer level will be (1) AND the Pressurizerheaterswillbe (2)

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

2. OFF B. 1. DECREASING 2.ON C. 1. INCREASING

2. OFF D. 1. INCREASING 2.ON Tuesday, July 13, 2010 Page 84 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 30 2530 neraI Discussion - On the Pressurizer Level Master Controller, located on the NV CHARGING FLOW CONTROL Graphic in DCS, the LI (Limit Increase) and

LD (Limit Decrease) buttons are used to set a minimum limit LM for automatic charging flow to ensure seal injection flow to the NC Pumps is maintained. There is an LM setpoint window and also an LM bargraph displayed on the Level Master controller. The limit is set in gallons per minute. The normal setting is 35 gpm. This function is bypassed when the Pressurizer Level Master Controller or the SLIMs for NV-238 is placed in MANUAL. This function is also bypassed when the SLIMs for NV-238 is placed in L-MANUAL. This limit value is set up per OP/l(2)/A16200/OOlA (Chemical and Volume Control System Letdown) Enc. 4.1. In the event PZR Level decreases to 17%, valves NV1A, NV2A, NV457A, NV458A and NV35A are automatically closed. This isolates letdown to prevent further loss of inventory and minimize the possibility of uncovering the heaters. At the same time all PZR Heater groups are de-energized to protect them from overheating should they become uncovered. An Annunciator Alarm, PZR LO LEVEL HTRS OFF & LETDN SECURED, alerts the operator of the low level condition. Another feature which will isolate letdown and de-energize the pressurizer heaters is charging flow lowering to <20 gpm for> 20 seconds. With this question, the changing flow is lowered to 18 GPM which would result in a L/D isolation. Approximately 12 GPM will still be leaving the NC system via NCP seal leakoff so with 18 GPM total charging, PZR level will be increasing and PZR heaters will be off. - Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part (1) is plausible if the applicant fails to realize that letdown is isolated or concludes that NCP seal leakoff is greater than the current charging flow. Part (2) is correct and therefore plausible. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Part (1) is plausible if the applicant fails to realize that letdown is isolated or concludes that NCP seal leakoff is greater than the current charging flow. (2) is plausible because the heaters do not de-energize due to PZR low level until level reaches 17%. If the applicant fails to recall that aters will be off due to the low flow condition associated with charging this answer is plausible. Answer C Discussion CORRECT: See explanation above. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: Part (1) is correct and therefore plausible. Part (2) is This answer is plausible if the applicant does not recall that in addition to the letdown isolation when charging flow decreases to less than 20 GPM for 20 seconds the Pressurizer heaters are de-energized as well. Basis for meeting the KA - The Pressurizer is part of the RCS. Any malfunction that effects Pressuirzer level effects RCS inventory and any malfunction that effects Pressurizer pressure effects RCS pressure. Since these malfunctions/operations affect both Pressurizer pressure and level, RCS pressure and inventory are both effected. Therefore, the KA is matched. Basis for Hi Cog -- This is a higher cognitive level question because it require more than one mental step. First the applicant must analyze the given condition to determine the status of the LCS and the potential consequences of the initial conditions. The applicant must then recall from memory the protective features which can be affected by operating the level control system in the configuration given and determine which protective actions are going to occur and in what order. - -- -- Basis for SRO only - Job Level - Cognitive Level - QuestionType Question Source RO Comprehension NEW Wednesday, July 14, 2010 Page 85 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 30 - 2530 C Development References Student References Provided arning Objective: i) PS-ILE-DCS #17

References:

1) Lesson Plan OP-MC-PS-ILE-DCS Sections 2.4.1 & 2.5.1 SYSO1 1 K3.02 Pressurizer Level Control System (PZR LCS)

Knowledge of the effect that a loss or malflrnction of the PZR LCS will have on the following: (CFR: 41.7/45.6) RCS -9 Comments: - Remarks/Status Proposed revision for 2010 NRC Q-30. Revision appoved RFA 07/06/10. Tuesday, July 13, 2010 Page 86 of 294

Question 30 High Miss Question Proposed Replacement

References:

From Lesson Plan OP-MC-PS-ILE-DCS Section 2.4.1: When the Soft Control or the SLIMs for NV-238 is placed in Manual or the SLIMs is taken to L-MANUAL the Pressurizer Level Master Controller is swapped to Manual also by DCS. However, when the Soft Control or the SLIMs for NV-238 is returned to AUTO the operator must place the Pressurizer Level Master Controller back in AUTO. Objective #7 On the Pressurizer Level Master Controller, located on the NV CHARGING FLOW

CONTROL Graphic in DCS, the LI (Limit Increase) and LD (Limit Decrease) buttons are used to set a minimum limit LM for automatic charging flow to ensure seal injection flow to the NC Pumps is maintained. There is an LM setpoint window and also an LM bargraph displayed on the Level Master controller. The limit is set in gallons per minute. The normal setting is 35 gpm. This function is bypassed when the Pressurizer Level Master Controller or the SLIMs for NV-238 is placed in MANUAL. This function is also bypassed when the SLIMs for NV-238 is placed in L-MANUAL. This limit value is set up per OP/i (2)/A/6200/OOIA (Chemical and Volume Control System Letdown) Enc. 4.1. Objective #8 When in MANUAL, the output of the controller sets a fixed position for NV-238. Increasing the output causes NV-238 to open, while decreasing the output causes NV-238 to close. Objective #4 2.4.2 NV-238 SLIMs Station This SLIMs station is used to control the position of NV-238. In AUTO, it compares the output of the Level Master to Selected Charging Flow (which is developed using a Median Select Algorithm with three charging flow inputs) to position the valve for needed charging flow. In MANUAL or L-MANUAL, UP/DOWN push-button arrowheads are used to position the valve. When the Soft Control or the SLIMs is taken to MANUAL or the SLIMs is taken to L MANUAL the Pressurizer Master Level Controller is swapped to MANUAL also by DCS. However, when the Soft Control or the SLIMs for NV-238 is returned to AUTO the operator must place the Pressurizer Level Master Controller back in AUTO. Objective #4 2.4.3 PD Pump SLIMs Station

This station is used to control the speed of the PD Pump. The Controller will be a MANUAL only controller. The UP/DOWN arrowhead push-buttons are used to adjust speed. If the AUTO pushbutton is depressed the LED on the AUTO pushbutton will illuminate and immediately return to the MANUAL pushbutton LED illuminating. From Lesson Plan OP-MC-PS-ILE-DCS Section 2.5.1: 2.5 Control Functions 2.5.1 PZRLowLevel Objective #9 In the event PZR Level decreases to 17%, valves NVIA, NV2A, NV457A, NV458A and NV35A are automatically closed. This isolates letdown to prevent further loss of inventory and minimize the possibility of uncovering the heaters. At the same time all PZR Heater groups are de-energized to protect them from overheating should they become uncovered. An Annunciator Alarm, PZR LO LEVEL HTRS OFF & LETDN SECURED, alerts the operator of the low level condition. Another feature which will isolate letdown and de-energize the pressurizer heaters is charging flow lowering to <20 gpm for> 20 seconds. The Selected Charging flow signal is developed with a Median Select algorithm with input from three (3) transmitters measuring charging flow. The low charging flow signal is maintained for 15 seconds and then clears, therefore if Pressurizer Level is >17% the Pressurizer Heaters can be placed back into service even though charging flow may not have been restored. Objective #11 Once level has increased to greater than 17% all heater groups must be manually re energized and letdown can be re-established. This is accomplished by selecting MAN on A, B, and D Heater MAN/AUTO Selector Switch. This allows closing the 600V supply breaker from their control switches on MC-5. C Heater supply breaker is closed via the switch on MC-1 0. There is no MAN/AUTO switch for C Heater. NOTE: If a Safety Injection has occurred, the Safety Injection signal and the sequencers must be reset in order to close the A & B heater breakers. 2.5.2 High Level Deviation Objective #9 If level should increase to greater than 5% above program level an Annunciator alarm, PZR HI LEVEL DEV CONTROL, is generated and the back-up heaters come on. This is done so that the subcooled water which has just surged into the PZR can be heated to saturation temperature. This will allow the water to flash to steam and avoid a pressure decrease as the level decreases to normal.

2.5.3 Low Level Deviation If level should decrease to less than 5% below program level an Annunciator alarm, PZR LO LEVEL DEVIATION, alerts the operator of the low level condition. 2.5.4 Hi Level Alarm If level should increase to 70% an annunciator alarm, PZR HI LEVEL, alerts the operator of the high level condition.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination SYSO14 24.31 Rod Position Indication System (RPIS)

QUESTION 31 r 2531 ( YS014 GENERIC (now1edge of annunciator alarms, indications, or response procedures. (CFR: 41.10 / 45.3) Unit 1 is operating at 100% RTP. The following indications are observed on the Digital Rod Position Indication (DRPI) system:

D-4 rod indication is RED
Associated rod group background is ORANGE
IAD-2 I D10 (RPI URGENT FAILURE) is LIT Which ONE (1) of the following describes the condition of rod D-4?

A. Rod D-4 is fully inserted. B. Rod D-4 is at half accuracy. C. Rod D-4 position cannot be determined. D. Rod D-4 is greater than 231 steps withdrawn. Tuesday, July 13, 2010 Page 87 of 294

_______ _______ _____ ______ ______ FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 31

   .,eneral Discussion iiioijiications are exhibited with a Data A and Data B failure:

a) Red position for indication for affected rods b) Red U above affected rods c) Red zero for affected rods position d) Red RB light e) Red Urgent alarm f) Orange background for affected banks g) Yellow Data Failure alarm (A and B) h) Yellow deviation alarm i) RPI Non-Urgent Annunciator j) RPI Urgent Annunciator Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible since the applicant may conclude from the indications that the rod is fully inserted and the indication is valid based on the given conditions. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: It is plausible for the applicant to conclude that the rod is at half accuracy due to a Data A OR Data B failure Answer C Discussion -- -

                                                                                                                               -______

[ERRECT: See explanation above. Answer D Discussion - INCORRECT: See explanation above. LAUSIBLE: This answer is plausible if the applicant concludes that a rod which is >231 steps withdrawn gives an RPI urgent failure. DRPI is Jt capabile of monitoring a rod greater than 231 step withdrawn and it would be reasonable for the applicant to conclude that this condition would result in an urgent failure which would be consistent with_any_other condition where DRPI could not determine actual rod_position. Basis for meeting the KA The KA is matched because the applicant must understand the meaning of numerous DRPI system alarms and their impact on the operation of the DRPI system. Basis for Hi Cog This is a higher cognitive level question. The applicant must recall what each DRPI alarm means with regards to the operation of the system. The applicant must then analyze from the multiple alarms given in the initial conditions the overall impact on the DRPI system. Since the question requires multiple mental steps to arrive at the correct answer, this is a higher cognitive level_question.________________ Basis for SRO only , Job Level , Cognitive Level , QuestionType, Question Source RO Comprehension BANK CNS 2008 RO Audit Retake Exam (Q 1857) Development References udent References Provided - - Learning Objectives:

1) IC-EDA#7&8

References:

1) Lesson Plan OP-MC-IC-EDA Section 3.2.1 SYSO14 2.4.3 1 Rod Position Indication System (RPIS)

S0l4 GENERIC now1edge of annunciator alarms, indications, or response procedures. (CFR: 41.10 / 45.3) Tuesday, July 13, 2010 Page 88 of 294

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2010 MNS SRO NRC Examination QUESTION 31 401-9 Comments: - emarksIStatus 401-9 Comments: I think the answer is obvious as written. Can you reduce the stem indications to make it less obvious? D is NP because of the 3rd bullet (Red RB is indicated for rod D-4) Replace D Resolution I Comments: Revised question and removed third bullet. Didnt really see how we could replace D with anything that was more plausible. Removing third bullet gives D plausibility. See attached file for ysed copy of question. Tuesday, July 13, 2010 Page 89 of 294

Question 31

References:

From Lesson Plan OP-MC-lC-EDA Section 3.2.1: DRPI Urgent Alarm Objective #7,8 Refer to Drawing 7.10, D.R.P.l Display DataA+ B Failure (Urgent). DRPI Urgent Alarm caused by Data A Failure and Data B Failure on a rod P-8:

The best calculated position indicated immediately below rod P-8 alpha-numaric designator would indicate a red 0.
The failure status line would indicate a red U above rod P-8 bar graph.
The rods bar graph would turn red and indicate rod height of 0.
The background color for Control Bank C would turn orange.
A red RB would be indicated on the rod bottom status line.
The system status line would indicate a yellow DATA A FAILURE, yellow DATA B FAILURE, red URGENT ALARM, and since the other rods in this bank are> 12 steps withdrawn a yellow DEVIATION> 12 STEPS.

Refer to Drawing 7.11, D.R.P.l Display Rod Deviation (Urgent) DRPI Urgent Alarm caused by an actual deviation of 12 steps on a rod P-8:

The background color for Control Bank C would turn orange.
The system status line would indicate a red URGENT ALARM and a yellow DEVIATION

        > 12 STEPS condition.

Refer to Drawing 7.12, D.R.P.I Display Gray Codes Disagree (Urgent) DRPI Urgent Alarm caused by gray codes not in agreement on rod P-8 with the result the best calculated position is 12 steps or more from other rods in the bank.

The best calculated position indicated immediately below rod P-8 alpha-numaric designator would indicate a average of Data A and Data B.
The rods bar graph would turn yellow.
The background color for Control Bank C would turn orange
The system status line would indicate a red URGENT ALARM and a yellow DEVIATION

       > 12 STEPS condition.

Note that if the gray codes not in agreement resulted in an averaged position within 12 steps of the other rods, there would be no deviation or urgent indications. The only indication would be the rod would turn yellow with a RPI Non-Urgent Annunciator. An example of this scenario is when leads for Data A and Data B are rolled, and rods are withdrawn. DRPI sees the B coil made first, knows this is a disagreement and intermittently turns the rod yellow (until the A coil is made), but the indicated position never gets 12 steps from the other rods, so no deviation and no urgent alarm.

_______ FOR REVIEW ONLY DO NOT DISTRIBUTE - 2008 CNS RO Audit Retake Examina QUESTION 57 1857 r SYSO14 Knowledge of RPIS design feature(s) andlor interlock(s) which provide for the following: (CFR: 41.5 / 45.7)r]Rod bottom lights K4.03 Unit 1 is operating at 100% power. Given the following indications on the Digital Rod Position Indication (DRPI) system:

Associated bank background is orange
D-4 rod indication is red
Red RB is indicated for rod D-4
1AD-2, DuO RPI URGENT FAILURE is alarming Which one of the following describes the condition of rod D-4?

A. Rod D-4 is at half accuracy B. Rod D-4 at greater than 231 steps withdrawn C. Rod D-4 is fully inserted D. Rod D-4 position cannot be determined Monday, February 22, 2010 Page 113 of 151

_________________ _______ FOR REVIEW ONLY DO NOT DISTRIBUTE - 2008 CNS RO Audit Retake Examina QUESTION 57 iss General Discussion The following indications are exhibited with a Data A and Data B failure: a) Red position for indication for affected rods b) Red U above affected rods c) Red zero for affected rods position d) Red RB light e) Red Urgent alarm 0 Orange background for affected banks g) Yellow Data Failure alarm (A and B) h) Yellow deviation alarm

1) RPI Non-Urgent Annunciator

 ) RPI Urgent Annunciator Answer A Discussion Plausible: The student may believe the rod is at half accuracy due to a Data A OR Data B failure Answer B Discussion Plausible: The student may believe that rod is >231 steps withdrawn gives an RPI urgent failure.

Answer C Discussion Plausible: The student may believe the rod is fully inserted and the indication is valid based on the given conditions Answer D Discussion Correct: The indications given are for a Data A and Data B failure for rod D-4 Basis for meeting the KA Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 2007 Audit Examination#2 Q53 (Bank 53) Development References Student References Provided EDA KA KA_desc SYSO 14 Knowledge of RPIS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.5/ 45.7)ERod bottom lights K4.03 401-9 Comments: [arksIStatus Monday, February 22, 2010 Page 114 of 151

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 32 SYSO 15 K2.O 1 Nuclear Instrumentation System (NIS)

nowledge of bus power supplies to the following: (CFR: 41.7) JS channels, components, and interconnections Given the following conditions on Unit 1:

Unit is shutdown in MODE 6 for Refueling
While responding to a senes of alarms associated with the Nis the operator notices that the Instrument Power and Control Power lights on the PR N43 drawers are DARK Which ONE (1) of the following is the cause of these indications?

A. inverter I EV1A has tripped. B. The feeder breaker for panelboard 1 EKVB has tripped. C. Inverter 1 EVIC has tripped. D. The feeder breaker for panelboard 1 EKVD has tripped. Tuesday, July 13, 2010 Page 90 of 294

____________ ___________ ___________ ___________ ___________ ______ __________ ______ _____ _______ FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 32 12532 eneral Discussion - NIS Channel 3 (PR N43) is powered from IEKVC which is fed from Static Inverter Answer A Discussion iiCT: See explanation above. PLAUSIBLE: This answer is plausible because Static Inverter 1EVIA supplies panelboard IEVCA which powers NIS Channel 1 (N31, N35, andN41). Answer B Discussion INCORRECT: See explanation above. SIBLE: This answer is plausible because panelboard IEKVB provides power to NIS Channel 2 (N32, N36, and N42). Answer C Discussion CORRECT: See explanation above. Answer D Discussion above. PLAUSIBLE: This answer is plausible because panelboard_1EKVD provides power to NIS Channel IV (N44). Basis for meeting the KA KA is matched because the applicant must know the power supplies for all NIS channels to determine the correct answer. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Learning Objectives: EL-EPL #5 & 6

References:

1. Lesson Plan OP-MC-EL-EPL Section 1.2

2. Lesson Plan_OP-MC-IC-ENB Section 2.7 SYSO15 K2.01 Nuclear Instrumentation System (NIS)

Knowledge of bus power supplies to the following: (CFR: 41.7) NIS channels, components, and interconnections 401-9 Comments: Remarks/Status 40 1-9 Comments: No comment. Resolution / Comments: N/A Tuesday, July 13, 2010 Page 91 of 294

Question 32

References:

From Lesson Plan OP-MC-EL-EPL Section 1.2: 120 VAC Vital Instrumentation and Control Power System I Objective#5 I The 120 VAC Vital Instrumentation and Control Power System consist of four vital panelboards and four inverters to each unit. The four vital panelboards will normally receive powerthrough static inverters 1(2) EVIA, 1(2) EVIB, 1(2) EVIC and 1(2) EVID. A regulated power supply (1 KRP for Unit I and 2 KRP for Unit 2) is also provided, as an alternate power source, to allow uninterruptible manual power transfer to panelboards 1(2) EKVA, 1(2) EKVB, 1(2) EKVC, and 1(2) EKVD when an inverter is intentionally taken out-of-service. This system provides four independent channels for instrumentation and control power to both units (Unit 1 and 2). A Train loads are fed from channels A and C while the B Train loads are fed from channels B and D. Three of the four channels will ensure that the overall system functional capability is maintained, comparable to the original design standards for safe operation. However, a loss of any two of these channel sources will result in a shutdown of the respective unit. Objective # 6 The following is a listing of typical loads that are powered from the 120 VAC Distribution Centers:

NIS Channels 1 thru 4 Instrument Power
NIS Channels 1 thru 4 Control Power
SSPS Instrument Power
SSPS Control Power
FWST Channels I thru 4 Instrument Power
Containment Radiation Monitors Isolation Valves
Auxiliary Safeguard Cabinets Instrument Power
Post Accident Recorders
Post Accident Annunciators 1.0 COMPONENT DESCRIPTION 1.1. 125 VDC Vital Instrumentation and Control Power System Battery Chargers The two-unit station is provided with five battery chargers, designated EVCA, EVCB, EVCC, EVCD; and a spare battery charger, designated EVCS, which can be used to replace a charger if required. These chargers, supplied by SCI (Solid state Controls Incorporated), are 500 ampere chargers with a charging capability of 500-625 amps, however, we have them current limited at 525 amperes.

From Lesson Plan OP-MC-IC-ENB Section 2.7: 2.7 Power Supplies NIS Channel I EKVA NIS Channel II EKVB NIS Channel Ill EKVC NIS Channel IV EKVD Wide Range Train A EKVA Wide Range Train B EKVD 3.0 SYSTEM OPERATION 3.1 Normal Operation 3.1.1 Operating Procedures The Excore Nuclear Instrumentation System provides the operator with neutron flux indication for all modes of operations. During each reactor startup, a healthy skepticism (N concerning the validity of power indications is warranted, particularly following a V refueling outage. Changes in plant equipment or conditions, along with a strong desire to return the plant to full operation, may influence personnel to accept less than complete explanations for discrepant indications. For example, excessive electrical generation for the nuclear power indicated (a symptom of miscalibrated nuclear instruments) has been attributed to factors such as: cold circulating water temperature, expected efficiency improvements, and changes in core design or instrumentation.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 33 253 SYSO 16 K4.O 1 Non-Nuclear Instrumentation System (NNIS)

nowledge of NNIS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) rteading of NNIS channel values outside control room Which ONE (1) of the following sets of indications can be read outside the Main Control Room on BOTH the Auxiliary Shutdown Panel (ASP) AND the Safe Shutdown Facility (SSF) Control Panel? A. SR Neutron Flux AND SIG WR Levels B. SR Neutron Flux AND Pressurizer Level C. Incore Thermocouples AND SIG WR Levels D. Incore Thermocouples AND Pressurizer Level Tuesday, July 13, 2010 Page 92 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 33 eneraI Discussion Pressurizer level and SR Neutron Flux can be read outside the Main Control Room on both the ASP and SSF Control Panels. Answer A Discussion - --______________________________ INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because SR Neutron Flux can be read on the both the ASP and the SSF. SO Wide Range level can be read on the SSF but not on the ASP. Answer B Discussion CORRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because S/G WR Levels and Incore Thermocouples call both be read on the SSF Control Panel. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because Incore Thermocouples can be read on the SSF Control Panel and Pressurizer level can be read on both the ASP and SSF. Basis for meeting the KA The KA is matched because the applicant must recall all indications (both Nuclear and Non-Nuclear indications) available at the SSF and_A Basis for Hi Cog Basis for SRO only Job Level Cognitive Level LQ5ti0Type Question Source RO Memory NEW Development References Student References Provided Learning Objectives:

1. CP-ASP #2

2. CP-AD #8

References:

1. Lesson Plan OP-MC-CP-AD Section 2.1

2. Lesson Plan OP-MC-CP-ASP Section 2.1 -

SYSOI6 K4.01 Non-Nuclear Instrumentation System (NNIS)

Knowledge of NNIS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) Reading of NNIS channel values outside control room 401-9 Comments: RemarkslStatus 401-9 Comments: Consider adding an additional indication to increase LOD. A,A (none good) A,B (B good only) B,B (All good) B,A (B good only Resolution / Comments: Developed a revised question with two answers per distracter. If Tuesday, July 13, 2010 Page 93 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 33 2533 revised question is used the distracter analysis will need to be revised. See attached file for revised copy of question. Tuesday, July 13, 2010 Page 94 of 294

Question 33

References:

From Lesson Plan OP-MC-CP-ASP Section 2.1: 1.0 COMPONENT DESCRIPTION 1.1. Panel Indications (Refer to Drawings 7.1, 7.2, 7.3, & 7.4) 1.1.1. Temperature indications (all temperature indication is continuous)

Reactor Coolant System Wide Range Hot Leg Temperature (0-700°F) Loop D Hot Leg
Reactor Coolant System Wide Range Cold Leg Temperature (0-700) Loop D Cold Leg
Regenerative Heat Exchanger Letdown Temperature (100-600)
A & B ND Pump Discharge Temperature (50-400°F)
A, B, C, & D ND to Cold Leg Temperatures (50-400°F) 1.1.2. Pressure Indications (all pressure indication is continuous)
Wide Range Reactor Coolant System Pressure (0-3000 psig)
Narrow Range Reactor Coolant System Pressure (PZR Press) (1700-2500 psig)
Letdown Pressure (0-600 psig) 1.1.3. Level Indications (continuous)
Channel 1 Pressurizer Level (0-100%)

1 .1 .4. Flow Indications (continuous)

Letdown Flow (0-200 gpm) 1 .1 .5. Power Indication (continuous)
SR Nuclear Flux (101 - 5 cps, separately detected, not part of the NIS) i0

12. Manual Loaders on the Panel 1.2.1. NV-459 (Variable L/D Orifice Outlet Flow Control)

It is used to throttle letdown flow rate when initiating letdown. Throttling prevents thermal shock of the letdown piping by allowing the operator to slowly initiate letdown. Per procedures, excess letdown is established if normal letdown is not in service. 1.2.2. NV-21A (NV Spray to PZR Isol) It is used to control NC System pressure if the normal spray valves are unavailable or not functioning properly (Note: the normal spray valves should be operating in Auto, and no control of them on the ASP). Its used on the ASP for NC System pressure control during cooldown. Letdown must be in service before this valve can be used. This is to ensure the z\T between the Pressurizer Temperature and Spray Water is less than 320°F, which aids in preventing thermally shocking the spray nozzle. When NV-21 is being used, valves NV-13B and 16A must be closed (Normal and Alternate Charging) which allows the operator to maintain a more constant letdown and charging flow balance.

From Lesson Plan OP-MC-CP-AD Section 2.1: The pump is driven by an induction motor powered from the standby shutdown power supply. Control switches for the pumps and various isolation valves are located on the SSF Control Panel. A filter is provided downstream of the pump to collect any particulate matter larger than 5 microns that could cause damage to the reactor coolant pump seals. Filter differential pressure is indicated locally. Since the makeup pumps deliver a constant flowrate to the Reactor Coolant System, it may become necessary to remove excess water to maintain Pressurizer level 60 80%. - Solenoid operated, reactor vessel head vent valves (NC272 & 273) are powered by the Standby Shutdown system to allow discharge of water to the Pressurizer Relief Tank (PRT). Controls for these valves are located on the SSF Control Panel. A flowpath for the Standby M/U Pump is provided by opening NV842AC and NV849AC. These valves will close on a Phase A (Si) signal if they are being powered from their normal power supply (EMXA-4). Once control is swapped to the SSF and EMXA-4 is swapped to its alternate power supply (MCC SMXG) the valves will no longer close on a Phase A (Si) signal. Pressurizer level is indicated on the SSF Control Panel. 1.2.3. Temperature Indication can be monred from the SSF Control Panel to monitor vLondItIos A multi-conductor cable that is connected on the side of the control panel must be relocated in order to view the thermocouple readings. The highest reading Core Exit Thermocouple is used to determine subcooling. Indication is also provided for the Incore reference junction temperature deviation. This temperature deviation indication is used to obtain a corrected Core Exit Thermocouple value to be used in determining subcooling. Indication is also provided for Loop A and D WR Cold Leg temperatures.

1.2.4. Pressure Control In order to prevent steam bubble formation in the reactor vessel, primary pressure must be maintained above saturation pressure at the core exit temperature. A sub-group of Back-Up Heater Group D (7O kW) is powered from the SSF electrical distribution system and can be controlled from the SSF Control Panel. The heaters are energized as necessary to maintain subcooling if pressure decreases. This ensures the steam bubble stays in the Pressurizer. The heaters have a LOCAL/REMOTE switch and a control switch. The LOCAL position bypasses all AUTO and Control Room functions. The Pressurizer Spray valves can also be controlled from the SSF Control Panel. The spray valves have open/close switches which are used to ensure that the spray valves remain closed (gives a hard closed signal). The normal position for this switch is the closed position. This switch is only functional when controlling (via EMXA-4 swap) from SSF. Reactor Coolant System wide range pressure indication is provided on the SSF Control Panel. 1.2.5. Flux Indication WR Neutron Flux Indication is provided on the SSF Control Panel. Indication is provided from 1 01 CPS up to 10 CPS. 1.3. Secondary System Control m Generator Wide Range level indication is provided on the SSF ContJ These level indicators are calibrated fo. hot conditions since the design of the SSF is to iiiaintain Hot StandA The TD CA Pump will auto start if 1/1 WR level transmitter indicates 72% on 2/4 S/Gs. A step in the body of AP-24 Loss of Plant Control due to Fire or Sabotage will have the operator manually start the TD CA pump prior to leaving the control room and a step in AP-24 Enc. 1 will place SA-48ABC in the open position at the SSF. Procedurally the TDCA flow will be controlled based on the availability of the controls in this order: control room, the CA pump room or locally in the doghouses. A steam supply is assured to the TD CA Pump on swap over to the SSF due to the MSIV and S/C PORV on C S/C failing closed. Feedwater is also assured to provide a heat sink due to the CA supply valves (CA 54AC and CA 66AC) from the TD CA Pump to B and A S/Cs failing as is (with a normal position of open) on swap over to the SSF. Feedwater is provided to C S/G from the TD CA Pump by verifying CA5OB (TD CA to S/C C Isol) open and securing the hand wheel clutch in the engaged position as directed by procedure. NOTE: The word disengaged in the next paragraph refers to the motor not the handwheel. Permanently installed step ladders were added in the basement of the doghouse near CA54AC and CA66AC. The motor operator clutch levers for CA38B, CA5OB, CA54AC,

and CA66AC have eyelets such that an eyebolt can be screwed into them to secure the lever in the disengaged position. The eyebolts are stored on the clutch lever plates using a short piece of small wire. Labels are attached with this wire which indicates that eyebolts are dedicated for use during certain SSF Events. A two position switch for SA-48ABC (A FWDT Steam Supply) is located on the SSF Control Panel to prevent continual cycling of the TD CA Pump. The two positions are: AUTO: SA-48ABC will open in response to an auto start signal. OPEN: Seals in SA-48ABC in the open position and bypasses the auto start signal. This switch will normally be maintained in the Auto position. It will be selected to open by Enclosure 1 of AP-24 to seal in the auto start signal to the TDCA pump. NOTE: This switch will only affect the SSF related solenoids.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 34 SYSO28 A2.O I Hydrogen Recombiner and Purge Control System (HRPS)

lalfunctions or operations on the HRPS; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of jose malfunctions or operations: (CFR: 41.5 /43.5 /45.3 / 45.13) Hydrogen recombiner power setting, determined by using plant data book Given the following on Unit 1:

A LOCA occurred 24 hours ago
The 1A H2 Recombiner was placed in service per EPI1IAI5000IG-1 Enclosure 4 (Placing H2 Recombiners In Service)
Containment pressure was 5 PSIG when the Recombiner was placed in service Current Conditions are as follows:
Containment pressure is 1 .5 PSIG Based on the conditions above the recombiner Power Setting was (1) when the recombiner was placed in service and should now be set to (2)

Which ONE (1) of the following completes the statement above? REFERENCE PROVIDED A. 1. 49.8KW

2. 45.3KW B. 1. 49.8KW

2. 45.8KW C. 1. 50.3KW

2. 45.3KW D. 1. 50.3KW

2. 45.8KW Tuesday, July 13, 2010 Page 95 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 34 253 General Discussion In the scenario given with this question the power setting for the H2 recombiner should initially be 49.8 KW based on the initial containment pressure. Based on the current containment pressure of 1.5 PSIG the power setting should be 45.8 KW. Initial Power setting - Pressure Factor, CP = 1.395 @ PSIG Reference Power = 3 5.670 KW Power Setting = CP x Reference Power Power Setting = 1.395 x 35.67 = 49.8 KW Current Power setting - Pressure Factor, CP = 1.285 @ 5 PSIG Reference Power = 3 5.670 KW Power Setting = CP x Reference Power [iower Setting = 1.285 x 35.67 = 45.8KW Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is correct. Part 2 is plausible if the applicant incorrectly reads the wrong pressure line on the graph. Incorrectly reading the graph is plausible since the major divisions are in 2 PSIG increments and the minor divisions are in 1/2 PSIG increments nswer B Discussion 3RRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: Both parts are plausible if the applicant incorrectly reads the wrong pressure line on the graph. Incorrectly reading the graph is plausible since the major divisions are in 2 PSIG increments and the minor divisions are in 1/2 PSIG increments. Answer D Discussion nJCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible if the applicant incorrectly reads the wrong pressure line on the graph. Incorrectly reading the graph is plausible since the major divisions are in2PSIG increments and the minor divisions are in 1/2 PSIG increments. Part 2 is correct. Basis for meeting the KA The KA is matched because the applicant is asked to determine the Power Setting for the recombiner under two different conditions. This requires the applicant to determine the Pressure Factor both conditions using the Power Correction Factor curve from the Plant Data Book and then calculate the correct Power Setting for each condition. Basis for Hi Cog This is a hi cog question because the applicant must read the Power Correction Factor graph from the Plant Data Book and use the information from the graph to calculate the correct power setting. Since this requires more than one mental step, it is a higher cognitive level question. Basis for SRO only Job Level T cognitive Level QuestionType Question Source RO Comprehension NEW ...evelopment References Student References Provided Lesson Plan OP-MC-CNT-VX Pg. 27 (Rev 23) U-i Data Book Curve 1.8 EP Generic Enc G-1 End. 4 OP-MC-CNT-VX Obj 7 Tuesday, July 13, 2010 Page 96 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE B

2010 MNS SRO NRC Examination QUESTION 34 2534 SYSO28 A2.0l Hydrogen Recombiner and Purge Control System (HRPS)

alfunctions or operations on the HRPS; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of nose malfunctions or operations: (CFR: 41.5/43.5/45.3 /45.13) Hydrogen recombiner power setting, determined by using plant data book Oi-9 Comments: RemarkslStatus 401-9 Comments: No comment. Resolution / Comments: Replaced should in stem of question with will based on

Geneal Comment from Lead Examiner. See attached copy of question for proposed_revision.

Tuesday, July 13, 2010 Page 97 of 294

From OP-MC-CNT-VX Pg. 27 (Rev 23) 3.2 Abnormal and Emergency Operation The control panels for the electrical Recombiners are located in the MG set rooms. The recombiner units are located in Containment such that they process a flow of Containment air containing hydrogen at a concentration typical of the lower containment compartments. This is because the Hydrogen Skimmer Fans discharge in the vicinity of the Recombiners and the Recombiners process that flow. There is no direct piping or duct connection between the Recombiners and the Hydrogen Skimmer Fans. The recombiner consists of a thermally insulated vertical metal duct with electric resistance metal-sheathed heaters provided to heat a continuous flow of Containment air (containing a low concentration of hydrogen), up to a temperature which is sufficient to cause a reaction between hydrogen and oxygen (between 1225°F and 1400°F). The recombiner is provided with an outer enclosure to keep out water coming from the Containment Spray System. The recombiner consists of an inlet preheater section, a heater-recombination section, and a mixing chamber. The warmed air passes through an orifice plate (should protect the recombiner from being overloaded from higher hydrogen concentrations up to 6.0%) and then enters the electric heater section where it is heated to approximately 1225-1400°F causing recombination to occur. Tests have verified that the recombination is not a catalytic surface effect associated with the heaters, but occurs due to the increased temperature of the process gases. Since the phenomenon is not a catalytic effect, poisoning of the unit as by fission products will not occur. The heater section consists of five assemblies of electric heaters stacked vertically. Each assembly contains individual heating elements. Operation of the unit is virtually unaffected in the event of a few individual heating elements failing to function properly. Objective #7 The recombiners are equipped with chromel-alumel thermocouples with a reference junction monitored with an RTD. Digital temperature meters are provided on the Hydrogen Recombiner Heater Temperature Monitor Panel ( refer to Drawing 7.3) located in the MG set rooms. The display is normally off but may be operated if desired by:

1) Power on

2) Unit will perform self diagnostics and Return: Command?,

3) Press AUTO key.

The unit will display sequentially the three thermocouples points (numbered 1, 2, 3) and the reference junction temperature (number 4). The value for the reference junction is not fixed and is used to perform reference junction compensation for the thermocouples inputs. The three thermocouples provide recombiner temperature indication during testing. Temperature indication is not required during a LOCA, so the thermocouples portion of the recombiners is non-safety related, and both trains are on the same panel.

Question 34

References:

NNLLL OBJECTIVE L L P P 0 OORSR ROO

7. Discuss the instrumentation and controls associated with the X X X X X Hydrogen Recombiners, to include:

       . Temperature readout
       . Power adjust potentiometer
       . Power out meter
       . Power out switch
       . Power available light.

8. Discuss the instrumentation associated with the Hydrogen X X X X X Analyzer Concentration Monitors.

9. Evaluate plant parameters to determine any abnormal system x x x x x conditions that may exist.

10. Given a limit and/or precaution associated with an Operating X X X X X Procedure, discuss its basis and applicability.

11 Concerning the Technical Specifications related to the VX System:

      . Given the LCO title, state the LCO (including any         X X X COLR values) and applicability.
      . For any LCOs that have action required within one        x x x hour, state the action.
      . Given a set of parameter values or system conditions,     x x x determine if any Tech. Spec. LCOs is (are) not met and any action(s) required within one hour.
      . Given a set of parameter values or system conditions and the appropriate Tech Spec, determine required action(s)
      . Discus the bases for a given Tech. Spec. LCO or

Safety Limit SRO ONLY

From OP-MC-CNT-VX Pg. 27 (Rev 23) 3.2 Abnormal and Emergency Operation The control panels for the electrical Recombiners are located in the MG set rooms. The recombiner units are located in Containment such that they process a flow of Containment air containing hydrogen at a concentration typical of the lower containment compartments. This is because the Hydrogen Skimmer Fans discharge in the vicinity of the Recombiners and the Recombiners process that flow. There is no direct piping or duct connection between the Recombiners and the Hydrogen Skimmer Fans. The recombiner consists of a thermally insulated vertical metal duct with electric resistance metal-sheathed heaters provided to heat a continuous flow of Containment air (containing a low concentration of hydrogen), up to a temperature which is sufficient to cause a reaction between hydrogen and oxygen (between 1225°F and 1400°F). The recombiner is provided with an outer enclosure to keep out water coming from the Containment Spray System. The recombiner consists of an in let preheater section, a heater-recombination section, and a mixing chamber. The warmed air passes through an orifice plate (should protect the recombiner from being overloaded from higher hydrogen concentrations up to 6.0%) and then enters the electric heater section where it is heated to approximately 1225-1400°F causing recombination to occur. Tests have verified that the recombination is not a catalytic surface effect associated with the heaters, but occurs due to the increased temperature of the process gases. Since the phenomenon is not a catalytic effect, poisoning of the unit as by fission products will not occur. The heater section consists of five assemblies of electric heaters stacked vertically. Each assembly contains individual heating elements. Operation of the unit is virtually unaffected in the event of a few individual heating elements failing to function properly. Objective #7 The recombiners are equipped with chromel-alumel thermocouples with a reference junction monitored with an RTD. Digital temperature meters are provided on the Hydrogen Recombiner Heater Temperature Monitor Panel ( refer to Drawing 7.3) located in the MG set rooms. The display is normally off but may be operated if desired by:

1) Power on

2) Unit will perform self diagnostics and Return: Command?,

3) Press AUTO key.

The unit will display sequentially the three thermocouples points (numbered 1, 2, 3) and the reference junction temperature (number 4). The value for the reference junction is not fixed and is used to perform reference junction compensation for the thermocouples inputs. The three thermocouples provide recombiner temperature indication during testing. Temperature indication is not required during a LOCA, so the thermocouples portion of the recombiners is non-safety related, and both trains are on the same panel.

MNS GENERIC ENCLOSURES PAGE NO. EPI11N5ODO.G-i Enclosure $ Page 1 of 8

UNIT 1 Placing H2 Recombiners In Service R25 e. ArIIcN...Exp:CTED a!flDnEE ;EEPcNsE I:T camnz:

1. Select one train of 112 Recombiner to be placed in service:

   . To start 1A F-b Recombiner, GO TO Step 2.

OR

To start lB Ft Recombiner, GOTO Step5.

2. Determine 1A H2 Recombiner power setting as follows:

a. Determine PRESSURE FACTOR, CP from Data Book Curve 1.8.

b. Multiply A REFERENCE POWER ksted on Data Book Curie 1.8 by PRESSURE FACTOR. CP to determine 1 A Hydrogen Recombiner Pover Setting.

la PIFZPEUE ;0tcE:;cZEEUfl ZAD, :D :we:

c. RecordiAPOWER SETTING

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 35 SYS033 Al .02 Spent Fuel

                       -             Pool Cooling System (SFPCS)

( \bility to predict and/or monitor changes in parameters (to prevent exceeding design limits) Jperating the controls including: (CFR: 41.5 / 45.5) associated with Spent Fuel Pool Cooling System Radiation monitoring systems Given the following events and conditions associated with the Unit 1 SFP:

A Lo-Lo alarm is received for OAC point M1A0004 (SFP Level)
The operators read (-)2.1 ft SFP level and steady on the main control board
The operating KF pump has tripped
An NEO reports a large leak in the auxiliary building but the leak has now slowed to a trickle For the event described above the leak must be associated with the KF pump (1) piping and (2) would be utilized to monitor increasing radiation levels associated with the loss of SFP level.

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

2. IEMF-42 (U-i Spent Fuel Bldg Vent)

B. 1. discharge

2. 1EMF-17 (Spent Fuel Bldg Refuel Brdg)

C. 1. suction

2. 1EMF-42 (U-i Spent Fuel Bldg Vent)

D. 1. suction

2. 1EMF-17 (Spent Fuel Bldg Refuel Brdg)

Tuesday, July 13, 2010 Page 98 of 294

_____ _____ ______ _____ ____ ______ ______ FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 35 2535 General Discussion In the scenario described in the stem of this question, the indication provided would be consistent with a SFP cooling system leak associated with the discharge piping. The design of the piping is such that the hole drilled in the discharge piping located 2 feet below the normal level (Indication of 0 feet) act as a siphon breaker. With the pump tripped, this type of leak should slow to a trickle once level goes below this value. IEMF-17 is an area monitor located on the refueling bridge and would be the most direct indication of any increase in rad levels associated with the falling SFP level. IEMF-42 uses a beta gas detector which monitors the SFP ventilation rad levels. However, 1EMF-42 is designed to detect fuel failure based on the release of fission product gases. IEMF-42 is located in the ventilation ducting in another building from the SFP and is shielded from background radiation. For the scenario_described, there would be no effect on 1EMF-42 indication. . Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part (1) is correct and therefore plausible. Part (2) is plausible because it does monitor radiation levels associated with the SFP building ventilation system and if the applicant misinterprets the indicated level to be low enough to cause extreme radiation level this would beareasonable answer.

                                                                                                                                             -       I Answer B Discussion                                     
                                                                                                                 -

CORRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: Part (1) is plausible if the applicant confuses the siphon breaker location to be on the suction piping verses the discharge piping. Part (2) is plausible because it does monitor radiation levels associated with the SFP building ventilation system and if the applicant isinterprets the indicated level to be low enough to cause extreme radiation level this would beareasonable answer. swer D Discussion CORRECT: See explanation above. PLAUSIBLE: Part (1) is plausible if the applicant confuses the siphon breaker location to be on the suction piping verses the discharge piping. Part (2) is correct and therefore plausible. Basis for meeting the KA There is no direct correlation between the ability to monitor Radiation Monitor System parameters to prevent exceeding design limits associated with the Spent Fuel Pool Cooling System. However, for this particular question the applicant is asked to evaluate a given set of conditions and predict the minimum design SFP level which would be expected if leak developed on the discharge piping for the Spent Fuel Pool cooling pump. Additionally, the applicant is asked to identify which EMF could be used to verify the presence of a leak and that the leak has stopped. For example, in addition to the fact that SFP level has stopped decreasing. the Operator could use IEMF-l7 as rad monitor indication would initially increase due to lowering SFP level and then stop increasing when the leak stopsL Basis for Hi Cog .-

This is a hi cog question because it involves a level of analysis of given situation, apply system knowledge and solve a problem. Basis for SRO only______ Job Level Cognitive Level QuestionType Question Source RO Comprehension J BANK Bank MNS FHKFNO1 eveIopment References - Student References Provided

   ;son Plan OP-MC-FH-KF Page 27 (Rev 30)

OP-MC-FH-KFObj. .

                                                                                                                    -____

SYSO33 A 1.02 Spent Fuel Pool Cooling System (SFPCS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with Spent Fuel Pool Cooling System Wednesday, July 14, 2010 Page 99 of 294

_______ FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 35 ss operating the controls including: (CFR: 41.5 / 45.5) idiation monitoring systems 401-9 Comments: RemarkslStatus 401-9 Comments: The distractor analysis said IEMF-42 will have little to no effect. If A is marginally correct then it can be arguably correct. Therefore, 2 potentially correct answers exist. This must be re-evaluated. This Q is U until resolved due to 2 possible correct answers. Resolution I Comments: The discussion should have stated that 1EMF-42 will have no effect instead oflittle to no effect. This event would be dealt with via entry into AP-41 (Loss of Spent Fuel Pool Cooling or Level). An alarm on IEMF-17 is one of the symptoms that prompts entry into AP-41. There is plausibility for IEMF-42 in that an alarm on this monitor would prompt entry into AP-25 (Spent Fuel Damage). However, IEMF-42 is a beta gas monitor and will only respond if there is damage to the fuel in the SFP. Revised the discussion and distracter analysis for A2 and C2. See attached file for proposed changes to the discussion and distracter_analysis. Tuesday, July 13,2010 Page 100 of 294

Question 35

References:

OP-MC-FH-KF Obj. 12 OBJECTIVES NNLLL OBJECTIVE L L P P 0 OORSR ROO 1 State the purpose of the Spent Fuel Pool Cooling System. X X X X 2 Draw a simplified diagram of the Spent Fuel Pool Cooling X X X X System (including all major components) per Training Drawing 7.1, Spent Fuel Pool Cooling System Simplified.

3 State the flowrates through each of the following flowpaths: X X X X

    . Spent Fuel Pool Cooling Loop
   . Spent Fuel Pool Purification Loop
   . Spent Fuel Pool Skimmer Loop 4   List the sources of makeup to the Spent Fuel Pool Cooling     X X X X System; including the source grade (i.e., borated, non-borated demineralized, and non-borated lake water).

5 Explain the conditions which would require assured makeup, X X X X X from the Nuclear Service Water System, to the Spent Fuel Pool Cooling System. 6 List the power supply for the following Spent Fuel Pool X X X X Cooling System Pumps (Unit 1 and Unit 2):

   . KF Pump(s)
   . KF Skimmer Pump(s) 7  Describe the controls, indications, and/or alarms, associated x x x with Spent Fuel Pool Cooling System operation, located within the Control Room.

8 Describe how the KF Pump motor(s) is cooled during system X X X X operation. 9 State the cooling medium for the Spent Fuel Pool Cooling X X X X System Heat Exchanger(s). io Describe the controls, indications, and/or alarms, associated x x x x with Spent Fuel Pool Cooling System operation, located outside the Control Room.

From Lesson Plan OP-MC-FH-KF Page 27 (Rev 30) The Spent Fuel Pool stores fuel assemblies approximately 33 feet 4 inches below the fuel pool operating deck with approximately 25 feet of borated water above the top of each fuel assembly. Objective U 7 Control Room Indication is provided for Spent Fuel Level and Temperature. (Refer to Training Drawing 7.3, Spent Fuel Pool Control Room Indication.) In each of the Spent Fuel Pools and refueling cavities there is an Aztec Level Gauge. The angle iron pointing out into the water is at elevation 771 4%. This is the normal design level and corresponds to 0 on the gauge in the Control Room. Each step on the side edge of the gauge is two inches. (Refer to picture 7.5) 2.2 Spent Fuel Pool Cooling Pumps r Objective # 7 Two Spent Fuel Pool Cooling Pumps (KF Pumps) are provided for each Unit. The controls and indications, associated with Spent Fuel Pool Cooling Pump operation, located on the Main Control Board (MC-11), consist of the following:

START / STOP Control Switch These momentary START / STOP pushbuttons allow the operator to START and STOP the pump, as desired.

During a Blackout the KF Pump(s) will initially lose power (load shed) but receive a manual start permissive when Load Group 9 is loaded onto the bus. During a Safety Injection Signal, the KF Pump(s) running prior to SI will continue to run. The KF Pump(s) not running, prior to SI, will receive a manual start permissive when Load Group 9 is loaded onto the bus. Any KF Pump(s) running or manually started, while the SI Signal is present, cannot be stopped until the SI Signal is RESET.

ON / OFF (Red / Green) Indicating Lights These ON 10FF (Red I Green) indicating lights are mounted on the START / STOP Control Switch and provide indication when the KF Pump breaker is CLOSED (ON) or OPEN (OFF).

Typical flow through the heat exchanger and purification loop is 2500 gpm combined (approximately 2200 gpm through Hx and 300 gpm through purification). Each pump is designed for 3050 gpm and limited by procedure to 2900 gpm, and each takes suction from the Spent Fuel Pool, four feet below pool level, and discharge back into the Spent Fuel Pool, six feet above the fuel assemblies. Holes drilled into the Spent Fuel Pool Discharge Header act as a vacuum breaker and limit siphon draining to two feet below normal Spent Fuel Pool level.

From Lesson Plan OP-MC-FH-KF Page 53 (Rev 30) Abnormal Operating Procedure AP/1(2)/A/5500/25, Spent Fuel Damage, is provided to identify operator actions required during a spent fuel damage event. Actions are defined for spent fuel damage inside Containment or within the Spent Fuel Pool. This procedure has only a single Case and the Symptoms are:

EMF-36, Unit Vent High Gas Radiation Alarm (Process Monitor)
EMF-38, Containment High Particulate Radiation Alarm (Process Monitor)
EMF-39, Containment High Gas Radiation Alarm (Process Monitor)
EMF-40, Containment High Iodine Radiation Alarm (Process Monitor)
EMF-42, Fuel Handling High Gas Radiation Alarm (Process Monitor)
EMF-16, Containment Refueling Bridge Alarm (Area Monitor)
EMF-17, Spent Fuel Building Bridge Alarm (Area Monitor)
Gas bubbles originating from the damaged assembly(ies).
Visual evidence of damage with potential of radioactive release(s).

Subsequent operator action(s) will first determine the damaged fuel location. The area affected (Containment or the Spent Fuel Pool) must be evacuated and isolated. Those personnel evacuated must be assembled for accountability while remote action(s) are performed to further secure the event to ON-SITE. In addition, the event must be classified and implementafion of the Emergency Plan initiated, if required.

Parent Question Question 375 FHKFNO1 FHKFNO1 1 Pt(s) Unit 1 is operating at 100% power when the OAC registers a low spent fuel pool level alarm. Given the following events and conditions:

The operators read -2.1 ft SEP level and steady on the main control board.
The operating KE pump has tripped.
An NLO reports a large leak in the auxiliary building.
Normal SFP makeup is not available.

Which one of the following statements correctly describes the corrective action for this event? A. Find and isolate the leak on the KF discharge piping. B. Find and isolate the leak on the KF suction piping. C. Initiate assured makeup due a leak on the discharge piping. D. Initiate assured makeup due to a leak on the suction piping.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 36 SYSO35 K1.O1 - Steam Generator System (S/GS) ( Knowledge of the physical connections and/or cause-effect relationships between the S/GS and the following systems: (CFR: 41.2 to 4L9 I 45.7 to 45.8) MFWIAFW systems Given the following conditions on Unit 1:

A unit shutdown is in progress
Operators have blocked the CA Auto-Start signal
At 0200 both Main Feed Pumps trip Given the following plant conditions and times:

Time Condition 0200 0205 0210 0215 0220 Tave(°F) 551 552 552 553 554 NC Press. (PSIG) 1951 1953 1958 1951 1957 NRSGA(%) 24 16 25 18 10 NRSGB(%) 26 18 22 14 9 NRSGC(%) 28 20 26 13 8 NR SC D (%) 23 15 16 19 9 Which ONE (1) of the following lists the EARLIEST time that the Turbine Driven CA pump would have automatically started? A. 0205 B. 0210 C. 0215 D. 0220 Tuesday, July 13, 2010 Page 101 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 36 General Discussion The Turbine Driven CA Pump will auto-start when NR level on any two SGs decreases to less than 17%. For the conditions given, the Turbine Driven CA Pump will auto-start at 0205. Answer A Discussion CORRECT: See explanation above. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant concludes that NC system pressure must increase above the P-li setpoint to automatically unblock the CA Auto-Start Defeat AND also believes that only one SG less than 17% is required to generate a TD CA pump auto-start. However, the Auto-Start Defeat only applies to the MD CA pumps, not the TD CA pump. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant concludes that NC system pressure must increase above the P-li setpoint to automatically unblock the CA Auto-Start Defeat since two of the SG NR levels are less than the 17% level required for a TD CA pump auto-start. However, the Auto-Start Defeat only applies to the MD CA pumps, not the TD CA pump. Answer D Discussion ______ INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant concludes that NC system pressure must increase above the P-il setpoint to automatically unblock the CA Auto-Start Defeat AND that all four SG NR levels must be less than 17% to generate a TD CA pump auto-start signal. However, the Auto-Start Defeat only applies to the MD CA pumps, not the TD CA pump. Basis for meeting the KA The KA is matched because the applicant must understand the cause-effect relationship between SG level and the auto-start signals generated for e AFW (CA) system. 1 4asis for Hi Cog This is a higher cognitive level question because the applicant must associate multiple pieces of information to arrive at the correct answer. First, the applicant must recall from memory the coincidence and setpoint for the TD CA pump start and the effect of the CA Auto-Start Defeat on CA pump operation (MD CA pumps only). Then, the applicant must compare the information given in the table to the setpoint and coincidence recalled from memory to determine the correct answer. Since this question requires more than one mental step to arrive at the correct answer, it is a higher cognitive level_question. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK MNS Exam Bank Question #CFCANO 10 Development References ient References Provided Lesson Plan OP-MC-CF-CA Section 2.2 Learning Objective OP-MC-CF-CA #4 SYSO35 Kl.0l Steam Generator System (S/GS)

Knowledge of the physical connections and/or cause-effect relationships between the S/GS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8) MFW/AFW systems J4O19 Comments: RemarkslStatus - 40 1-9 Comments: No comment. Resolution / Comments: Tuesday, July 13, 2010 Page 102 of 294

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2010 MNS SRO NRC Examination QUESTION 36 2536 N/A Tuesday, July 13, 2010 Page 103 of 294

Parent Question: CFCANO1 0 1 Pt During a plant shutdown on Unit 1, the operators have blocked the CA auto-start signal by depressing the auto-start defeat switch. A subsequent loss of both main feedwater pumps occurred at 0200. Given the following plant conditions at the times listed: Time Condition 0200 0205 0210 0215 0220

1) Tave(°F) 551 552 552 553 554

2) NCS pressure (psig) 1951 1953 1958 1951 1957

3) NRSGA(%) 24 16 25 18 10

4) NR SC B (%) 26 18 22 14 9

5) NRSGC(%) 28 20 26 13 8

6) NRSGD(%) 23 15 16 19 9 What time would the Turbine Driven CA Pump start automatically?

A. 0205 B. 0210 C. 0215 D. 0220 Answer 938 A Objective 4

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 37 2537 SYSO45 K5.23 Main Turbine Generator (MT/G) System

nowledge of the operational implications of the following concepts as the apply to the MT/B System: (CFR: 41.5/45.7) elationship between rod control and RCS boron concentration during T/G load increases Given the following conditions on Unit 1:

Reactor Power is currently being increased from 55% to 90% RTP at 3%/hr following a Refueling Outage

1. How is the withdrawal of control rods affected?

2. What changes (if any) to NCS boron concentration will be required?

REFERENCE PROVIDED A. 1. NOT restricted

2. Dilution is required.

B. 1. NOT restricted

2. Dilution is NOT required.

C. 1. Restricted

2. Dilution is required.

D. 1. Restricted

2. Dilution is NOT required.

Tuesday, July 13, 2010 Page 104 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 37 C

.aeneraI Discussion With conditions given, the plant is above the conditioned power level therefore above 40% RTP, rod withdrawal is restricted to less than 3 steps per hour per the rod maneuvering limit guidance in the U-i Data book. This restriction on Rod movements would result in additional dilutions required to compensate for the negative reactivity associated with power defect during the power escalation.

Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible if the applicant does not recall the effect of unconditioned fuel on rod movement. The applicant may conclude based on plant conditions that there is no restriction on control rod movement under the conditions given. Part 2 of the question is correct and therefore plausible. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible if the applicant does not recall the effect of unconditioned fuel on rod movement. The applicant may conclude based on plant conditions that there is no restriction on control rod movement under the conditions given. Part 2 is plausible if the applicant confuses the effect of Xenon in the scenario described in the stem. On a power escalation after a runback, Xenon would burning out and adding positive reactivity. Answer C Discussion CORRECT: See explanation above. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is correct and therefore plausible. rt 2 is plausible if the applicant confuses the effect of Xenon in the scenario described in the stem. On a power escalation after a runback, enon would burning out and adding positive reactivity. Basis for meeting the KA This K/A is matched because the question is relating the effect of a T/G load increase during an initial power escalation with unconditioned fuel. The applicant must evaluate how this would affect the relationship between Rod control and RCS boron concentration due to the limitations imposed on rod movement. Basis for Hi Cog This is a hi cog question because it involves a level of analysis of given situation, apply system knowledge and solve a problem of what both would be the effect and how the conditions given in the stem would affect operation. It also requires more than one mental step to arrive at the correct answer and is therefore a higher cognitive level question. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Lesson Plan OP-MC-CTH-CP (Rev 11) Pages 135, 171, 173 &175 Data Book Sect. 1.3 Enc. 4.3 OP-MC-CTH-CP Obj: 29 SYSO45 K5 .23 Main Turbine Generator (MT/G) System

Knowledge of the operational implications of the following concepts as the apply to the MT/B System: (CFR: 41.5 / 45.7) elationship between rod control and RCS boron concentration during T/G load increases

)1-9 Comments:                                                                     marksIStatus 401-9 Comments:

Of the 4 bullets: I think you can delete all but the second bullet. Tuesday, July 13, 2010 Page 105 of 294

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2010 MNS SRO NRC Examination QUESTION__37 2537 In Stem 1. add at 3% per hr In Cl and Dl state Control rod withdrawal is restricted. In Al and Bi cap the word NOT Change B2 and D2 to No dilution will be required because Xenon burnout will compensate for the power defect Resolution I Comments: Deleted last two bullets. You need the first two bullets as a minimum. Made the rest of changes as recommended by Lead Examiner. See attached file for proposed revision to question. Tuesday, July 13, 2010 Page 106 of 294

Question 37

References:

-;_ E OORSR ROO 26 Given a set of plant parameters and/or system conditions, X X X X associated with the recovery of a misaligned / dropped rod, determine the appropriate recovery limits. CTHCPO26 27 Given a set of plant parameters or system conditions, X X X X associated with the recovery of a misaligned / dropped rod, discuss the basis for the appropriate recovery limits. CTHCPO27 28 Discuss the basis for the Fuel Maneuvering Limits. X X X X CTHCPO28 29 Given the Fuel Maneuvering Limits, evaluate a given set o X X X X plant conditions and determine the allowable loading I rod hdrawal rates. CTI-1CP029 30 Concerning the Technical Specifications related to Control Bank Insertion Limits, AFD, QPTR, and RCS Pressure, Temperature, and Flow DNB Limits:

   . Given the LCO title, state the LCO (including any COLR values) and applicability.
   . State the REQUIRED ACTION(s) and COMPLETION                        X X X TIME for action(s) with completion times of one hour or less.
   . Given a set of parameter values or system conditions,              X X X determine if any Technical Specification LCO(s) is (are) not met and any action(s) required within one hour.
   . Given a set of plant parameters or system conditions and           X X X the appropriate Technical Specification(s), determine the REQUIRED ACTION(s) and COMPLETION TIME(s).
   . Discuss the bases for a given Technical Specification              X X X LCO.

CTHCPO3O

From Lesson Plan OP-MC-CTH-CP Pg. 135 (Rev 11) POWER LEVEL Increasing reactor power (steam demand) results in two changes, one direct and one indirect, which affect power distribution: Redistribution (Direct Effect) Control Rod Movement (Indirect Effect) The first effect is the result of the variation in core AT with Reactor Power Level. As power is increased with turbine load, the core AT will rise from almost OoF at zero power to 58oF at full power. As a result the moderator in the upper portions of the core becomes progressively warmer and less dense relative to the bottom. The increasing density difference will force power toward the bottom of the core as evidenced by AFD becoming more negative The strength of this

redistribution effect is dependent on the value of the moderator temperature coefficient (MTC). At BOC when the MTC is small and negative, the redistribution effect is small. As the MTC becomes more negative with Burnup, the redistribution effect will become more pronounced. The second (indirect effect) is caused by the movement of control rods necessary to compensate, in part, for the power defect. As power is raised positive reactivity must be added in order to compensate for the negative reactivity associated with the power defect. Any rod withdrawal will tend to allow more power to be produced in the upper portions of the core, resulting in a tendency for AFD to become more positive. In practice the control rods are moved as necessary to offset the redistribution effect thereby maintaining a relatively constant axial power distribution. This is accomplished by coordinating reactor coolant boron concentration with rod position as necessary to maintain AFD on Target during the power escalation.

From Lesson Plan OP-MC-CTH-CP Pg. 171,173 & 175 (Rev 11) 3.3 Fuel Maneuvering Limits Objective # 28 The Fuel Maneuvering Limits apply to power increases ONLY. These maneuvering limits are tied to REACTOR POWER not Turbine or Generator Power. These limits are based on limiting or preventing PCI (Pellet-Clad Interaction). The primary concern is centered around previously used fuel and not new (fresh) fuel. Handling burned fuel, coupled with the fact that burned fuel has experienced fuel pellet cracking, can result in the movement of small pellet fragments. A gradual controlled increase in power will allow pellet and cladding expansion to somewhat equalize, as the fuel and cladding heat up. Objective # 29 I

Fuel Maneuvering Limits POWER RAMP RESTRICTIONS Recommended ramp 3% / hour, CPL 100% FP No restrictions <40%RTP Initial Startup foHowin fueling Shutdown or Recommended ramp YES 40% 100% FP Max Ramp = 3% Cold Shutdown wher assemblies handl: step change, 4% in 1 hour, 7% in 2 hours, and 10% in 3 hours. OPERATING CATEGORY 1: 100% F.P. for 72 umulative hours out of any Highest Power Level sustained fo 4oPeratinPeriodatPowe 72 cumulative hours (consecutive NO__* non-consecutive) out of any D 7 da operating period at power. YES 4 OPERATING CATEGORY 2: Highest Power Level sustained for 72 cumulative hours (consecutive n-consecutive) out of the preceding Recommended ramp = 3% I hour, CPL 100%

FP Max Ramp = 3% step change, 4% in 1 hour, 7% in 2 hours, wer40%F.P. YES YES and 10% in 3 hours. NO CPL 100% FP Recommended ramp = 3% / hour, CPL 100% FP Max

Ramp = 3% step change, 4% in 1 hour, 7% in 2 hours, and 10% in 3 hours.

.itial Startup following ng Shutdown or foil Shutdown where f ssemblieed. YES <with C R s INSERTED;> NO Power Levels. YES

ED Withdraw CRs as far as practical and keep CR mot to a minimum limiting to 3 steps! hour above 50% F. RESTRICTIONS on F.P.-NO4dIt50%FP YES I Withdrawal To Max. Of 3 steps! hour while maintaining 3% ! hour Max. Power Ramp. NO Determine Condition A. A is Max. withdrawal position and corresponding power level occurring during Startup. Upon withdrawal Restrcit withdrawal to beyond condition A Max Rate of 3 steps! hour while power level and CR maintaining 3% ! hour position Max. Ramp. For withdrawal beyond Condition A restrict withdrawal to Max Rate of 3 steps! hour and 3% I hour Max. Ramp Determine Condition A. A is Max. withdrawal NO RESTRICTIONS on C. R. withdrawal to Conditon A, YES position and corresponding NECESSARY Power Level,v and C.R. 4 el occurring ie NO For withdrawal beyond Condition A. Restrct withdrawal to a Max. Rate of 3 Steps! hour and 3%! hour Max. Ramp.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 38 SYSO71 K4.06 Waste Gas Disposal System (WGDS)

nowledge of design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) ( Jampling and monitoring of waste gas release tanks Given the following plant conditions:

Waste Gas Decay Tank A is aligned for planned release
Waste Gas Decay Tank E is also mistakenly aligned for release while in service
EMF-50 (L) Waste Gas Discharge is not detecting release activity Which ONE (1) of the following would be the result if the release exceeds expected activity levels?

A. The release is monitored by 2EMF-36(L) (Unit 2 Unit Vent Gas). However, no automatic termination will occur. B. The release is monitored by 2EMF-36(L) which will automatically terminate the release if a Trip 2 alarm is reached. C. The release is monitored by 1 EMF-36(L) (Unit 1 Unit Vent Gas). However, no automatic termination will occur. D. The release is monitored by I EMF-36(L) which will automatically terminate the release if a Trip 2 alarm is reached. Tuesday, July 13, 2010 Page 107 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 38 2538 .eneraI Discussion In the conditions given, a misalignment has resulted in EMF-50 not being aligned to properly monitor the WGDT release. The WG release is monitored by two EMFs, the primary is EMF-50 and the secondary is the U-i Unit Vent gaseous monitor 1EMF-36L. Activity detected resulting in a Trip 2 on either one of these monitors will result in a termination of the release due to the resulting auto closure of WG-i60. The release will still be monitored and auto termination is still functional. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: First part is plausible if the applicant confuses 1EMF-36L and 2EMF-36L. Second part is plausible if the applicant does not recall that the release can be terminated by the Waste Discharge monitor (EMF-50) or the Unit Vent Monitor (1EMF-36L). Answer B Discussion - INCORRECT: See explanation above. PLAUSIBLE: First part is plausible if the applicant confuses 1EMF-36L and 2EMF-36L. Second part is plausible because the Unit Vent Monitor will terminate the release on a Trip 2 alarm. However, it is 1EMF-36L instead of 2EMF-36L. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: First part is plausible if the applicant confuses 1EMF-36L and 2EMF-36L. Second part is plausible if the applicant does not recall that the release can be terminated by the Waste Discharge monitor (EMF-50) or the Unit Vent Monitor (1EMF-36L). nswer D Discussion ORRECT: See explanation above. Basis for meeting the KA This K/A is matched because the waste gas decay tank is being released and the applicant is being asked about both the design features (U-i release path) and interlocks (Action for EMF Trip 2) with regard to the monitoring capability associated with the release and the tank. Basis for Hi Cog Basis for SRO only Job_Leveif Cognitive Level QuestionType Question Source RO Memory BANK MNS Q WEWGNO3 Development References Student References Provided Lesson Plan OP-MC-WE-WG Page 29 (Rev 12) I OP-MC-WE-WG Obj. 5 - SYSO71 K4.06 Waste Gas Disposal System (WGDS)

Knowledge of design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) Sampling and monitoring of waste gas release tanks 401-9 Comments: RemarkslStatus 401-9 Comments: Distrsactors A and B are NP because there is no case where an isolation will not occur without a malfunction. Replace A and B. This Q is U because of 2 NP distractors. Tuesday, July 13, 2010 Page 108 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 26 SYSO78 A4.O1 Instrument Air System (lAS)

( bility to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) cressure gauges Due to a leak on the VI system the following indications were observed:

1AD-12 Cl (VINS Lo Pressure) is LIT
OVIP-5090 (VINS Press) dropped to a lowest reading of 86 PSIG and is now 89 PSIG and increasing Which ONE (1) of the following describes automatic actions which have occurred as a result of the indicated pressure transient?

A. G and H VI Compressors auto-started ONLY. B. lVl-820 (VI to VS Supply) auto-closed ONLY. C. 1VI-820 auto-closed AND lVl-1812 (VI Dryer Bypass VIv) has auto-opened. D. G and H VI Compressors auto-started AND 1VI-820 (VI to VS Supply) auto-closed. Tuesday, July 13, 2010 Page 72 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 26 2526 0 eneral Discussion At a decreasing VI pressure of 90 PSIG the following actions occur: 1VI-820 (VI to VS Supply) Auto closes G and H Compressors (Diesel VI compressors) Auto Start If VI pressure continues to decrease to 85 PSIG, 1VI-1812 (VI Dryer Bypass) will OPEN. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because this action would have occurred but is not complete. Answer is incomplete and incorrect due to the ONLY designation. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because this action would have occurred but is not complete. Answer is not complete and is incorrect due to the ONLY designation. Answer C Discussion INCORRECT: See explanation above. - -__________ PLAUSIBLE: This answer is plausible because the applicant may conclude that 1VI-1812 actuates with the other components at 90 PSIG. The first part is correct. Answer D Discussion CORRECT: See explanation above. Basis for meeting the KA KA is matched because the candidate, given information obtained from monitoring a trend of VI pressure indications located in the control oom, what automatic actions have occurred associated with the Instrument Air system. asis for Hi Cog

 ,iis is an analysis level question because the applicant must evaluate a given set of plant conditions, must recall a setpoint from memory, and then compare the plant conditions to the recalled memory to eliminate distracters and determine if a set of automatic actions should have occurred.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK M1JS (Bank 2227) Development References Student References Provided Lesson Plan OP-MC-SS-VI Objective 7 Section 1.2.10 page 67 and Objective 2 Section 1.3.1 page 89 ARP for 1AD-12 Cl (VI/VS Low pressure) SYSO78 A4.0l Instrument Air System (lAS)

Ability to manually operate and/or monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) Pressure gauges 401-9 Comments: RemarkslStatus 401-9 Comments: No comment. Resolution / Comments: N/A Tuesday, July 13, 2010 Page 73 of 294

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2010 MNS SRO NRC Examination QUESTION 26 252 Tuesday, July 13, 2010 Page 74 of 294

Question 26

References:

NNLLL OBJECTIVE L L P P 0 OORSR R0O 6 Explain the control function associated with each of the X X X X following VI Air Compressor (A, B, and C) pushbuttons:

     . Start/Stop pushbutton
     . Reset pushbutton List the interlocks / trips associated with operation of the   x x x x x following plant air system components:
   . VI Air Compressors
   . Vl-820 (VI to VS Supply Valve)
   . VS Low Pressure Air Compressor
   . VBAir Compressor 8  Describe the following controls and/or indications associated  X  X  X  X with operation of VI Air Compressors D, E, and F:
   . On/Off switch and indication
   . Start/Stop pushbuttons
   . Pre-lube pump status
   . Acknowledge/Reset pushbutton 9  Describe how the following VI System components function to    X  X  X  X provide a continuous supply of clean dry air:
   . Service Building Air Receiver Tanks (and drains)
   . Air Dryers
  . Auxiliary Building Instrument Air Tanks 10  Explain each one of the following controls and br indications, X  X  X  X associated with the Breathing Air Compressors:
  . Start/Stop Pushbutton
  . Power ON Light
  . RUN Light
  . Discharge Air Over-Temperature Light
  . Rotor Oil Filter Service Light
  . Bearing Oil Filter Service Light
  . Air/Oil Separator Service Light
  . Service Air Filter AP Gauge
  . Purification Filter AP Gauge
  . Rotor Coolant Temperature Gauge
  . Discharge Air Pressure Gauge
  . Discharge Air Temperature Gauge 11 Describe normal operation of the Breathing Air                  X  X  X  X  X

I Compressor(s). I From Lesson Plan OP-MC-SS-Vl Pg 71 (Rev 33) The VI System normally supplies the Low Pressure VS System through control valve 1 VI-820. Controls and indication for 1 VI-820 are located at the VI Sequencer Control PaneL The valve control switch is a three position switch:

Close
Auto
Open Objective # 7 Indication provided at the VI Sequencer Control Panel consists of the following:
1VI-820 Close (green light)
lVl-820 Open (red light)

This valve is normally in the AUTO position and will automatically close should VI System Pressure decrease to <90 psig. Upon valve closure IVI-820 can be reopened once VI System Pressure has increased >90 psig by placing the valve to the OPEN position. After opening the valve lVl-820, the switch should be returned to the AUTO position. If not, the valve will reopen without operator action, after closure, as soon as pressure has increased above 90 psig. 1.2.13 VI System Air Dryers A, B, and C Objective # 9 VI Dryers A, B, and C (AMLOC-CHA Dryers) are fully automatic, desiccant-type air dryers designed to remove vaporous moisture from the Instrument Air System. Generally, two of the three desiccant air dryers (A, B, and C) are in-service while one remains in standby, ready and available for service when needed. Each in service dryer will alternately cycle air through one of the two desiccant chambers for moisture removal, while the other chamber is regenerated (removal of previously adsorbed moisture) and re-pressurized.

From Lesson Plan OP-MC-SS-Vl Pg 75 (Rev 33) Purge Dump Restrictor Closes during dump and limits gas flow to prevent fluidization by controlling the rate of depressurization. Opens fully during all other periods. Dryer System Bypass Valve 1VI-1 812 is installed between the Dryer System Manual Bypass Valves 1VI-093 and lVl-094. This valve is designed to fail open on a loss of power or loss of air. Valves lVl-093 and IVI-094 will be normally open while 1Vl1812 will be normally closed. A solenoid operator associated with valve 1VI-1812 is connected to pressure switch 0V1PS5381. The solenoid is set to vent the actuator upon receipt of a VI System Low Pressure si nal 85# . switch 0VlPS531iscbnnectedioth1nstrt 0V1PS5380, which curren y controls the dryer Purge Exhaust Isolation Va 838 IVI-1839 and lVl-1 840) which fail closed on a low pressure sig S5381 sends a signal to the REFLASH Panel such .that alrrii orffFI wjjjjpdjcatQa_V!yer Panel .-oule There is local indication of valve position, RESET and OVERRIDE capabilities provided at the Reflash Panel. By depressing RESET, 1VI-1 812 will close, and by depressing OVERRIDE, 1VI-1 812 can be manually opened. 1VI-1812 is designed to automatically open and bypass the VI Dryers in the event of sudden blockage of flow due to some dryer malfunction. The PRA Study identified VI Dryer malfunctions as a primary contributor to Loss of VI event probability. The manual bypass valves (1VI-093 and 1VI-094) cannot protect against sudden dryer flow blockage events (e. g. switching valve failure). A filter is installed at the inlet of 1VI-1812 to prevent the potential of substantial contamination of the normally dry VI headers with rust known to exist in the wet VI headers. Instruments and Their Basic Function The A, B, and C VI Dryers are equipped with a set of gauges to indicate inlet air pressure, outlet air pressure, purge flow, and chamber pressure. The gauges are provided to monitor system operation. The gauges on the chamber indicate which chamber is on-stream (the gauge on the off-stream chamber should indicate zero (0) PSIG). The gauges are also used to verify that the internal pressure has been completely vented to the atmosphere when servicing is required. All pressure gauges should indicate zero (0) PSIG before any service work is performed on the dryer. Additional instruments include:

Chamber pressure relief valves.

Provide chamber protection if high pressure should develop during dryer operation. Set to relieve at design pressure.

Chamber Pressure Sensors Set to sense the lack or presence of chamber pressure following repressurization or depressurization.

From Lesson Plan OP-MC-SS-VI Pg 67 (Rev 33) Objective # 4 The Diesel VI Compressors operate in two modes of operation. These modes are Automatic and Manual. In the Manual Mode of operation, an operator will start and run the compressor using controls on the compressor control panel located at the compressors themselves. For a manual start of the compressor to be accomplished, the following must be true:

The AUTO/OFF-RESET switch must be selected to the OFF-RESET position
The STARTIWARM-UPIRUN switch is in the WARM-UP Position
The HIGH/LOW switch is selected to the desired position (normally HIGH)

The operator then rotates the Engine Switch from the OFF position to the ON position and the compressor should start. Once the compressor has started and has warmed up, the operator can select the RUN position on the STARTIWARM UP/RUN selector switch to allow the compressor to load. If the operator is starting the compressor as directed from the Loss of Instrument Air System Abnormal Procedure, the AP directs the operator to leave the START/WARM UP/RUN switch in the RUN position to allow for immediate loading. The following is a set of conditions, which will allow the Diesel VI Compressors to automatically start:

The AUTOIOFF-RESET switch must be selected to AUTO
The STARTIWARM-UPIRUN switch is selected to RUN
The HIGH/LOW switch is selected to HIGH
The Latching Relay picks up The compressor will automatically start and load to the desired pressure.

Objective # 7 There are three signals, which will send an AUTO START signal to the Diesel Powered VI Compressors. These signals are:

Loss of VI header pressure as measured by OVIPS5O7O

   ** set at 90 psig decreasing
   + Compressor control can be regained when pressure increases above 95 psig

Loss of 3/3 KR flow to the D, E, and F VI Compressors
Loss of power to the VI Sequencer Panel (SKU#43) ISLXD/2SLXD-SMXU

MNS Bank Question 2227: Due to a leak on the VI system the Unit 1 OATC observes the following indications:

1AD-12 Cl (VINS Lo Pressure) is LIT
CVI P-5090 (VINS Press) dropped to a lowest reading of 86 PSIG and is now 89 PSIG and increasing Which ONE (1) of the following describes automatic actions which have occurred as a result of the indicated pressure transient?

A. G and H VI Compressors Auto Started ONLY B. IVI-820 (Vito VS Supply) Auto Closed ONLY C. 1VI-820 Auto Closed IVI-1812 (VI Dryer Bypass Vlv) has Auto Opened D. G and H VI Compressors Auto Started jj 1VI-820 (VI to VS Supply) Auto Closed ANSWER: D

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 27 SYSO78 K3.02 Instrument Air System (lAS)

(nowledge of the effect that a loss or malfunction of the lAS will have on the following: (CFR: 41.7 /45.6) Systems having pneumatic valves and controls Given the following:

Unit 1 is operating at 100% RTP when a loss of VI event occurs
AP-22 (Loss of VI) has been implemented
VI header pressure is 55 PSIG and decreasing Which ONE (1) of the following system effects would be the FIRST to require the crew to trip the reactor in accordance with AP-22?

A. Decreasing S/G levels B. Loss of RN supply to Containment C. Loss of NC pump seal leakoff to the VCT D. PZR level approaching the High Level Trip setpoint Tuesday, July 13, 2010 Page 75 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 27 Aneral Discussion The CF control valves use 0 60# valve operating air. Depending on the nature of the problem with VI and considering line losses, etc., these

valves could start failing at 70# or more VI pressure as indicated in the control room. The operating philosophy regarding loss of Main Feedwater at power is to trip the reactor. This will prevent challenging the Lo-Lo S!G automatic reactor trip and will result in better initial conditions at the time of the manual trip. If the CF valves were to get to less than 25% open (for 30 sec or more) on 3 out of 4 SIGs, an AMSAC could also be generated. For most scenarios, its likely the operator will have manually tripped the reactor_prior_to this_occurring. Answer A Discussion CORRECT: See explanation above. Answer B Discussion - INCORRECT: See explanation above. PLAUSIBLE: IRN-252B does fail closed which would result in a loss of NSW cooling to the U-l NCPs. This is a significant operational concern and left in this condition would result in the need to trip the reactor and secure the NCPs. It is therefore plausible but incorrect because this condition would not be an immediate threat. Answer C Discussion INCORRECT: See explanation above.

    .PLAUSLE: 1NV-34A fails open but if the applicant believes that the failure mode of this valve is closed this would result in a loss of D/P across the A NCP #1 seal and require an immediate reactor trip and pump shutdown.

Answer D Discussion - ____________ 6ORRECT: See explanation above. PLAUSIBLE: 1NV-238 does fail open which would result in maximum charging flow. This would represent a longer term operational concern but would eventually result is challenging the PZR high level trip setpoint and is therefore plausible. Basis for meeting the KA This KIA is address because the applicant must understand the effect of a Loss of VI will have on 4 different pneumatic valves and how this loss

,,.

( uld affect the systems containing these components.

      .sasis for  Hi Cog This is a hi cog question because it involves a level of analysis of given situation, apply system knowledge and predict an outcome.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK MNS Exam Bank Question AP22NO1 Development References Student References Provided 12 AP-22 (Rev 28) Page 8 AP-22 Bacdground Document Page 15 OP-MC-AP-22 Obj. 5 -_______________ SYSO78 K3.02 Instrument Air System (lAS)

Knowledge of the effect that a loss or malfunction of the lAS will have on the following: (CFR: 41.7 / 45.6) Systems having pneumatic valves and controls 401-9 Comments: RemarkslStatus 401-9 Comments: Distractor D is a long shot and I do not believe it is plausible especially since other, more pronounced reactor trip criteria exists. Consider replacing distractor D. D is NP. Tuesday, July 13, 2010 Page 76 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 27 252 -_____ olution I Comments: Distracter D is not the strongest distracter. However, it is possible and is therefore plausible. Would like to keep this one. Tuesday, July 13, 2010 Page 77 of 294

Question 27

References:

OP-MC-AP-22 Obj. 5 OBJECTIVES NNLLL OBJECTIVE L L P P 0 OORSR ROO Concerning AP111A155001022 (Loss of VI): X X X

       . State the purpose of the AP
       . Recognize the symptoms that would require implementation of the AP AP22001 2  Describe the mitigating strategies (major actions) contained in        X X X the procedure.

AP22002 3 Given scenarios describing accident events and plant X X X conditions, evaluate the basis for any caution, note, or step. AP22003 4 Given scenarios describing accident events and plant X X X conditions, evaluate conditions which require application of continuous action steps. AP22004 5 State the failure modes of the components listed in AP/22, X X X Enclosure 12 (Valve Failure Mode on Loss of Air).

From AP-22 Enclosure 12 pages 1 of 6 MNS LOSS OF VI FACE NO.

Enclosure 12-Page lot S 10 1121 R E JENH I Valve Failure Mode on Lass of Air

NOTE The va.;es Isted in this enclosure faii at ahous ai: pressures. SB valves:

a. The fol:oMn; SE valves fal closed:

               . lBB-lB(IA S:C Sowdown Cant Outside Isol Control

               . 165-25 (15 SIC Slowdown Cant Outside Isol Control,

               . 1DB-SB (IC SIC Slowdown Cant Outside Isol Conoli

               . 136-43 (1D S/C Blowoown Cont Outside sal Controi:

155-54.4 SIC BE Cont Inside Isol: I . 166-64 lB SIC SB Cant Inside Isol:

               . 136-74 C S/C SE Cont lnde Isol:

135-84 ID SIC BE Cont inside Isal)
1 BE-i 23 :1 A SC Slowdown Throttle Control.i I . 135-124 :15 SIC Bbwdown Throttle Control:

               . 135-125:10 SIC Blowcaim Throttle Control:

               . 155-126 lID SIC Slowdown Throttle Canto:.

2. CA valves:

a. The folLowing CA vaves fah open:

               . 104-604 (14 CA Pirrp DischTo 14 SiC Controll

104-564 (IA CA Pump Disch To 15 S3 Control:

               . 104445 (15 CA Pump Dlsch To IC S/C Control:

               . 104405 15 CA Pu np Disch To ID S/C C or.trol:

               . 104-6445 :1 TD CA Pdmp Disch To 14 S       G Control) 1

               . 104-5248 U1 TD CA Pump Discn To 15 SC ControI
          .      1044846 /W TD CA Pump Disch To 10 S/C Control:

               . 104-3645 jUl TD CA Pump Disch To 1D SIC Conuo::.

3. CF valves:

a. The followinc CF alves a I closec:

               . 1CF-32A5 14 SIC CF Contro!)

1CF-23A6 16 S/C CF Control:

               . 1OF-20A6 IC S/C CF Control

1CF-WAS çlD S.C CF Control
1CF-10446 .14 SIC CF Control 5ypass

              . 1CF-105A5 .15 SIC CF Control      Bypass:

               . 1OF-106A3 (10 S/C CF Control      Bypass

               . 1CF-10745 :1D S.C CF Control      Bypass..

From AP-22 Enclosure 12 pages 3 of 6 MNS LOSS OF VI PAGE NO. AP1Ai55DCI22 121 Enclosure 12-Page 3 of 6 UNIT 1 Valve Failure Mode on Loss of Air t R Ct.

8. NV valves:

a. The following NV va.es fail open:

            . 1N/-iSA (N.. Supply To D NC Loop lcd:

1NV-i3B (NV Supply To A NC Loop Isoli

1N/-34A :4 NC Pump Seal Return lsol)

            . 1N/-5O5 (B NC Pump Seal Return col:

            . 1NV-664 (C NC Pump Seal Return Isol:

            . 1N/-62B (0 NC Pump Seal Return lcd)

1N/-i24 (Letdown Pressure Control:

lNv-238 (Charging Lire Flow Conrol,

INV-241 UI Seal Water lnj Flow Control:

            . 1N.-267A :Bcric Ac:d To Elenoe Contrail.

h. The following NV .ahes fail to the CT posbon:

1NV-27B (Excess L:D I-h 0th 3-Way Cntrl:
1 274 LiD H Outlet 3Way Temp Cntrl:
I NVI 374 :NC Fhters Ott .3Way Cntrl (.

c. The following NV alQes fail closed:

           . 1NV-IA (NC Ut sol To Regen H<:
       .      1 NV-24 (NC LID lsol To Regen Hx:

           . 1N/-DlA (NV Spray To PZR lool)
       .      1N7-245 (C NC Loop To Es LID H. Isol)

1N/-DSB (C NC Loop To Exs LID H.: lcd)
1 N.-26B (U 1 Excess LID Hx Outlet Cntrl:
1 N/-354 (Variable LID Orifice Outlet Corn Isoli

1N39A (A NC Pump Standpipe Fill:

           . 1NV-SSB (B NC Pump Standpipe Fill

           . 1N/-71A (C NC Pump Standpipe Fill

           . 1NU-87B (0 NC Pump Standpipe Fl

           . 1NV-92A (NC Pumps Seal Gyp Return Hdr Isol:

1 NVi 21 (Ut ND Letdown ControL:
1NV-iG7A (VCT Vent To WO isol,

           . 1NV-1714 (BA B:enderTc VCT Inlet:
          . 1NV-I7SA (BA Blender to CT Outlet
          . 1NI-457A (45 GP1i LiD C)dfice Cutlet Cont Isol,

           . 1N/-458A (75 GPM LID Orifice Outlet Contlsoli
          . 1N7459 (Ut Variable LID Orifice Cutlet Flow Cntrli

           . 1N.-B4QA (Ui ND To Pzr Aux Spray Control.

From AP-22 Enclosure 12 pages 3 of 6

9. RE valves:

a. The foko/ding RE \alve fails closed:

1RF-6214 (Unit I RF Cant Outsice Itoh.

10. RN valves:

a. The foltowing RN cakes fail open:

1RNOSA (RN to A KC Hx Cot 1RN-103A iA NV Pump Cooler Sup Itoh 1RN-I14A (A. NI Pump Cooler Sip isoli 1RN-126A A NS Pump ESS AHU SLp loon 1RN-13DA (A ND Pump ESS AHU Sup sol.i 1RN-140A (A. KF Pump ESS AHU Sup sol: 1RN-19DB tRN To B KG H Contol: 1RN-204B 6 NI Pump Cooker Sup sol 1RN-215B i,B NI Pump Cooler Su lscl 1RN-2276 ,B NS Pump ESS A.HU Sup lsoli 1RN-2316 k 6 ND Pump ESS AHU Sup (sol 1RN-245B i,B KE Pump ESS AHU Sup Itoh.

h. The fol.owh; RN calves fail cosed:

1RN-21A (1A RN Strainer Backwash Automatic Supply Itoh 1RN-22A (1A RN Strainer Backwash Autoniatic Drain 1RN-25B :16 RN Strainer Backwasn Automatic Supply sal 1RN-26B (16 RN Strainer Backwash Automatic Drain 1RN-252B (RB Non Ess Sup Cart Outside ksol) 1RN-277B RB Non Ess Ret Cont Outside 1501:. ii. RVvaives:

a. The fokowin; RI aIves fail closed:

1R:-79A (UI VU AHUS RV Cant :Duzside Supply Hdr ltd:
1RV-1Q1A. (U 1 VU .AHUS RV Cant Inside Retuvn Hdr Isol:
1RV-60B (UI VU .AHUS Ri Cont Inside Supply Hdrlsoli
1R!-102B iU 1 U AHUS RV Corn Outside Return Hdr Itoh.

From AP-22 Page 8 of 121: a:r:obl.EXEE::s: RrSoNLE Es;oNss NOT ce:aztzr I II. iConnuedi

m. Control NC temperature as folos:

Throme ND flow.

NOTE

KC to ND H flow should be close to flow prio: to loss of VI, since t is normally controlled by motor operated valves.

KC to ND H.. flow incicatons fails low during a loss of VI. Aternate indications are available at the foIowing ocations. if needed:

14: 1KCFT-5670 aux bag. 733 +2, west of column MM-E4:
1B: 1KCFT56ED {aux bldg. 733 ÷4, west side of column JJ-55L

              . IF NC temperature is greater than 2JDF.THEN maintain KC flow to ND Hx greater than 2020 GPM.

Throtte KC Flow to ID Hx as required.

_12. FAT.4ji VI pressure is less than 70 PSlG THEN align B Train RN to SNSWP PER Enolo sure 7 (Aligning B Train RN to Pond). NOTE CF Control Valves will fa I closed on tow VI press jre. which may result in AMSAC actuation and Lo Lo Sf0 leve. 13 Check Sf0 levels - AT PROGRAMMED IF 510 levels are going down in an LEVEL. uncontrolled mannerr T perform the following:

a. Trp reactor.

b Contnue with this procedure as time allows.

c. GO TO EP,1[:50DD;E-D iReactor Trip or Safety lnjecton i.

AP-22 Background document Page 15 STEP 13: PURPOSE: Prompt the operators to watch S/G levels because the CF control valves fail closed on a loss of VI. If SIG levels cant be controlled, the Operator is directed to trip the reactor. DISCUSSION: The CF control valves use 0 60# valve operating air. Depending on the nature of the

problem with VI and considering line losses, etc., these valves could start failing at 70# or more VI pressure as indicated in the control room. The operating philosophy regarding loss of Main Feedwater at power is to trip the reactor. This will prevent challenging the Lo-Lo S/G automatic reactor trip and will result in better initial conditions at the time of the manual trip. Refer to PIP 2-M-87-0208 where a automatic reactor trip occurred 5 mm after loss of offsite power due to loss of VI to the CF valves. If the CF valves were to get to less than 25% open (for 30 sec or more) on 3 out of 4 SIGs, an AMSAC could also be generated. For most scenarios, its likely the operator will have manually tripped the reactor prior to this occurring.

REFERENCES:

PIP 2-M-87-0208

Parent Question AP22NOI: Question 6 AP22NO1 1 Pt Unit 1 is operating at 100% power when a loss of VI event occurs. AP/1/A15500/22 (Loss of VI) has been implemented. VI header pressure is 55 psig and going down. Which of the following conditions would initially jeopardize the plant and require the SRO to direct tripping the Unit 1 Reactor per AP/1/A/5500/22 (Loss of VI)? A. 1 NV-238 (Charging Line Flow Control) fails closed. B. 1CF-23AB (B S/G CF Control Vlv) fails closed. C. 1 RN-252B (RB Non Ess Sup Cont Outside IsoI) fails closed. D. I NV-34A (A NC Pump Return Isolation) fails closed. Answer 6 B

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 28 SYS 103 A4.04 Containment System

                  -
\bility to manually operate and/or monitor in the control room: (CFR: 41.7/ 45.5 to 45.8) base A and phase B resets Given the following conditions on Unit 1:

A LOCA has occurred inside Containment
Containment pressure is currently 3.5 PSIG Which ONE (1) of the following describes the MINIMUM steps required before KC can be restored to Containment?

A. Reset PhaseA B. Reset Phase B C. Reduce Containment pressure below 1.0 PSIG, reset Phase A D. Reduce Containment pressure below 3.0 PSIG, reset Phase B Tuesday, July 13, 2010 Page 78 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 28 2528: Jeneral Discussion Phase B actuation secures Component Cooling Water (KC) to the Reactor Coolant pumps, Nuclear Service Water (RN) to the Reactor Coolant Pump Motor Coolers, Containment Ventilation Cooling Water (RV) and Instrument Air (VI) to the containment. Phase B can be reset with signal still present, once resets are pushed, we regain control of valves that close on the PhaseB signal. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant does not recall which signal (Phase A or Phase B) closes the ContainmentKC valves. Answer B Discussion 2RRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: The answer if the applicant does not recall whether the KC valves are closed by a Phase A or Phase B signal. If the applicant concludes that the valves are closed by a Phase A signal it is reasonable to also conclude that Containment pressure must be reduced to less than 1.0 PSIG (where aPhaseAsignal would be initiated by the Hi Containment pressure SI) in order to reset the the Phase A signal. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible since the KC valves are closed by a Phase B signal and the signal must be reset to open the valves. It is reasonable for the applicant to conclude that the Hi-Hi Containment pressure signal seals in which would prevent resetting the Phase B signal unless Containment pressure is reduced to less than 3.0 PSIG. Basis for meeting the KA - -_____________ By demonstrating a knowledge of when the Phase B reset must be operated to regain control of equipment operated by the Phase B signal, the pplicant demonstrates the ability to operate Phase B resets from the Control Room. Therefore the KAis matched. asis for Hi Cog Basis for SRO only JobLevel Cognitive Level jQuestionType Question Source [_____ RO Memory BANK MNS Exam Bank ECCISENO4 Development References Student References Provided Learning Objective:

1) ECC-ISE#13

References:

1) Lesson Plan OP-MC-ECC-ISE Section 3.1 SYS 103 A4.04 Containment System

Ability to manually operate and/or monitor in the control room: (CFR: 41.7/45.5 to 45.8) Phase A and phase B resets 401-9 Comments: RemarkslStatus 40 1-9 Comments: Since Phase B is actuated (3.5 psig), it would appear that the KC valves are closed as stated. The distractor analysis for D appears to indicate that Cntmt pressure must be reduced to < 3 psig before phase B can be reset (because phase A is still in). In other words can one reset phase B without phase A being reset? Please re-verify this because the reference was not clear on this issue. Tuesday, July 13, 2010 Page 79 of 294

____ _________ FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 28 IsQisintil re-verified. Resolution / Comments: Phase A does not have to be reset in order to reset Phase B. And, Phase B can be reset with containment pressure greater than 3.0 psig. The wording in distracter D analysis is incorrect. Instead of prevent resetting the Phase A signal unless Containment pressure is reduced is should have said prevent resetting the Phase B signal unless Containment pressure is reduced. Changed analysis for distracter D accordingly. Tuesday, July 13, 2010 Page 80 of 294

Question 28

References:

From Lesson Plan OP-MC-ECC-ISE Section 3.1: Objective # 13 Phase B Containment Isolation is actuated by: 2/4 Hi Hi Containment Pressure > 3.0 psig on channels Manually /2 pushbuffons Phase B actuation secures Component Cooling Water (KC) to the Reactor Coolant pumps, Nuclear Service Water (RN) to the Reactor Coolant Pump Motor Coolers, Containment Ventilation Cooling Water (RV) and Instrument Air (VI) to the containment. Phase B can be reset with signal still present, once resets are pushed, we regain control of valves that close on the Phase B signal. Containment Ventilation Isolation (SHy is initiated by any of the following:

Safety Injection (Ss)
Manual Phase A (Si)
Manual NSIPhase B
Trip 2 alarm on EMF-38, 39, or 40 Containment Ventilation Isolation (SH) signal secures VQ and VP.

To Reset Containment Ventilation Isolation following a Safety Injection, Manual Phase A, or Manual Phase B, the Containment Ventilation (SH) Reset Pushbuttons must be depressed (can reset without resetting the initiating signal). To Reset Containment Ventilation following an EMF 38, 39, 40 Trip II, the EMF must be reset, then the Containment Ventilation Reset Pushbuttons must be depressed. NOTE: Resetting the SH signal will allow manual control of VQ valves. VQ valves do not have an auto function.

Annulus Ventilation System (VE) start maintains negative pressure in annulus. It is actuated automatically by a Hi Hi Containment pressure signal or manually by either depressing Manual NS/Phase B Pushbutton or placing VE (Annulus Ventilation) to ON. To reset the start signal we must reset the Phase B isolation, then, place VE (Annulus Ventilation) fan switch to Reset and place back in auto. If 2 Skimmer and Air Return Fan (VX) starts on a Hi Hi Containment Pressure (Sr) with CPCS or Manually by NS/Phase B pushbutton and CPCS after a 10 minute time delay.

Question 28 Parent Question: ECCISENO4 1 Pt Given the following conditions:

1) Containment pressure is 3.8 psig

2) Phase B containment isolation has occurred What are the minimum steps required to restore Component Cooling water to containment?

A. Restore KC to operation immediately B. Reset Phase B, restore KC to operation C. Reset SI, reset Phase B, restore KC to operation D. Reduce containment pressure below 3.5 psig, reset Phase B, restore KC to operation Answer 599 Answer B MISCINFO: RO&SRO SOURCE: BCH

REFERENCES:

OP-MC-ECC-ISE page 29 LESSON: OP-MC-ECC-ISE TASK: OBJECTIVE: 1.N.2 TIME: K/A: 022000K403 (3.6*/4.0*) DATE: 11/29/95

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 29 2529 SYSOO1 K6.13 Control Rod Drive System

nowledge of the effect of a loss or malfunction on the following CRDS components: (CFR: 41.7/45.7) ocation and operation of RPIS Give the following conditions on Unit 1:

The unit is in MODE 3 withdrawing S/D banks in preparation for startup
IAD-2 / D1O (RPI Urgent Alarm) Annunciator has just alarmed
DRPI and OAC RODS position indication for rod D-8 has been lost What is the FIRST action required by SLC 16.7.9 (Rod Position Indication System -

Shutdown)? A. Place rods in manual ONLY. B. Place rods in manual AND drive all rods in. C. Immediately open the reactor trip breakers. D. Restore rod position indication within 1 hour. Tuesday, July 13, 2010 Page 81 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 29 2529

,eneraI Discussion SLC 16.9.7 (Rod Position Indication System Shutdown) requires that at least one rod position indicator be operable and capable of
                                               -

determining the control rod position within + 12 steps for each rod not fully inserted. This SLC is applicable to Modes 3,4,5. In the situation given in this question. the unit is in Mode 3 in the process of withdrawing S/D Banks. If rod position is lost for any rod, Condition A requires that the Reactor Trip breakers be opened immediately. Answer A Discussion - INCORRECT: See explanation above. PLAUSIBLE: With any malfunction involving the control rods this would be the required action in -]4. It In this case given this is not the [ect action because it is not required by SLC 16.9.7. - Answer B Discussion - INCORRECT: See explanation above. PLAUSIBLE: If the applicant correctly remembers that a shutdown is required but confuses the required action with a one hour requirement. The verification of shutdown margin is consistent with almost every 1 hour action statement concerning rod alignment and position indication with the unit in Mode 1 or 2 therefore it would be plausible for the applicant to apply that requirement to this situation. Answer C Discussion RRECT: See explanation above. Answer 0 Discussion INCORRECT: See explanation above. PLAUSIBLE: This would satisfy the requirements of TS 3.1.4 (Rod Group alignment limit Action B. With One rod not within alignment limits, Action B.I requires the rod to be restored within alignment limits within 1 hour. The applicant may incorrectly apply the actions of this spec because with the rod position indication unavailable it would be impossible to prove that it was within alignment limits. Basis for meeting the KA dthough there is no physical cause/effect relationship between the RPIS and CRDS, for this particular instance, a malfunction has occurred in e RPIS and the effect on the CRDS is that operator action is required by SLC 16.7.9 to immediately de-energize the CRDS. Therefore, the KA .s matched. Basis for Hi Cog Basis for SRO only Level RO Cognitive Level Memory J QuestionType BANK Question Source Bank MNS ICEDARO 1 Development References - Student References Provided SLC 16.9.7 SYSOO1 K6.13 Control Rod Drive System

Knowledge of the effect of a loss or malfunction on the following CRDS components: (CFR: 4 1.7/45.7) Location and operation of RPIS 401-9 Comments: Remarks/Status 40 1-9 Comments: B is NP as written. Place in the stem what is the FIRST action and remove and do not move them from distractor B. E because distractor B is NP as written. Resolution / Comments: Revised question per Lead Examiners recommendation. Then Tuesday, July 13, 2010 Page 82 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 29 2529 rearranged distracters A and B for psychometrics mading B the new correct answer. If this is acceptable the distracter analysis will need to be reworked. See attached file for proped revision. Tuesday, July 13, 2010 Page 83 of 294

Question 29

References:

OP-MC-IC-EDA Obj. 10

10. Concerning the Technical Specifications related to the DRPI System:

Given the LCD title, state the LCD (including any CDLR values) and applicability.
For any LCDs that have action required within one hour, state x x x the action.
Given a set of parameter values or system conditions, determine if any Tech. Spec. LCDs is (are) not met and any action(s) required within one hour.
Discus the bases for a given Tech. Spec. LCD or Safety Limit

                               *SROONLY x x x x

From Selected Licensee Commitment 16.9.7 16.7 INSTRUMENTATION 16.7.9 Rod Position Indication System - Shutdown COMMITMENT One rod position indicator (excluding demand position indication) shall be OPERABLE and capable of determining the control rod position within +/-

12 steps for each shutdown or control rod not fully inserted. APPLICABILITY MODES 3, 4 and 5 with the reactor trip breakers in the closed position with rods not fully inserted and capable of withdrawal. NOTE For testing or trouble shooting, alternate methods may be used to ensure there is no possibility of rod motion. These methods are pulling fuses, sliding links in the rod control cabinets or removal of CRDM head cables. After one of these alternate methods is used, the reactor trip breakers may remain in the closed position. REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required Open the reactor trip breakers. Immediately rod position indicators inoperable.

Parent Question ICEDAROI Question 29 ICEDARO1 1 Pt(s) Unit #1 is in Mode 3. While withdrawing Shutdown Bank E

1 the DRPI rod position indication for Rod D-8 was lost at 96 steps. The Rod Position Indication (RPI) urgent failure annunciator, General Warning for D-8, and Rod Bottom Light for D-8 were received. OAC Program General 76 does not update for Rod D-8 when Bank E is moved. Select the action which must be taken by the operator: A. Immediately trip the reactor B. Place rods in manual and do not move them C. Continue the startup but do not enter Mode I D. Drive all rods in and verify shutdown margin within 1 hour Answer 29 A

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 30 SYSO1 1 K3.02 Pressurizer Level Control System (PZR LCS)

( nowledge of the effect that a loss or malfunction of the PZR LCS will have on the following: (CFR: 41.7 / 45.6) 1 { CS Given the following conditions on Unit 1:

The unit is at 100% RTP
All Pressurizer heaters are energized in MANUAL
The SLIM for I NV-238 (Charging Flow Control) has been placed in MANUAL due to a malfunction of the Pressurizer Level Master Controller
The OATC reduces the 1NV-238 SLIM output to reduce Pressurizer level
Charging Line Flow is inadvertently reduced to 18 GPM If the 1NV-238 controller output remains constant, after 5 minutes Pressurizer level will be (1) AND the Pressurizer heaters will be (2)

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

2. OFF

, B. 1. DECREASING 2.ON C. 1. INCREASING

2. OFF D. 1. INCREASING 2.ON Tuesday, July 13, 2010 Page 84 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 30 2530

      .neraI Discussion On the Pressurizer Level Master Controller, located on the NV CHARGING FLOW CONTROL Graphic in DCS, the LI (Limit Increase) and
                                                                     -

LD (Limit Decrease) buttons are used to set a minimum limit LM for automatic charging flow to ensure seal injection flow to the NC Pumps is maintained. There is an LM setpoint window and also an LM bargraph displayed on the Level Master controller. The limit is set in gallons per minute. The normal setting is 35 gpm. This function is bypassed when the Pressurizer Level Master Controller or the SLIMs for NV-238 is placed in MANUAL. This function is also bypassed when the SLIMs for NV-238 is placed in L-MANUAL. This limit value is set up per OP/l(2)/A16200/OO1A (Chemical and Volume Control System Letdown) Enc. 4.1. In the event PZR Level decreases to 17%, valves NV1A, NV2A, NV457A, NV458A and NV35A are automatically closed. This isolates letdown to prevent further loss of inventory and minimize the possibility of uncovering the heaters. At the same time all PZR Heater groups are dc-energized to protect them from overheating should they become uncovered. An Annunciator Alarm, PZR LO LEVEL HTRS OFF & LETDN SECURED, alerts the operator of the low level condition. Another feature which will isolate letdown and dc-energize the pressurizer heaters is charging flow lowering to <20 gpm for> 20 seconds. With this question, the changing flow is lowered to 18 GPM which would result in aL/D isolation. Approximately 12 GPM will still be leaving the NC system via NCP seal leakoff so with 18 GPM total charging, PZR level will be increasing and PZR heaters will be off. Answer A Discussion - -

INCORRECT: See explanation above. PLAUSIBLE: Part (1) is plausible if the applicant fails to realize that letdown is isolated or concludes that NCP seal leakoff is greater than the current charging flow. Part (2) is correct and therefore plausible. Answer B Discussion - INCORRECT: See explanation above. PLAUSIBLE: Part (1) is plausible if the applicant fails to realize that letdown is isolated or concludes that NCP seal leakoff is greater than the current charging flow. (

(2) is plausible because the heaters do not dc-energize due to PZR low level until level reaches 17%. If the applicant fails to recall that aters will be off due to the low flow condition associated with charging this answer is plausible. Answer C Discussion _.. - CORRECT: See explanation above. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: Part (1) is correct and therefore plausible. Part (2) is This answer is plausible if the applicant does not recall that in addition to the letdown isolation when charging flow decreases to less than 20 GPM for 20 seconds the Pressurizer heaters are dc-energized as well. Basis for meeting the KA The Pressurizer is part of the RCS. Any malfunction that effects Pressuirzer level effects RCS inventory and any malfunction that effects Pressurizer pressure effects RCS pressure. Since these malfunctions/operations affect both Pressurizer pressure and level, RCS pressure and inventory are both effected. Therefore, the KA is matched. Basis for Hi Cog - This is a higher cognitive level question because it require more than one mental step. First the applicant must analyze the given condition to determine the status of the LCS and the potential consequences of the initial conditions. The applicant must then recall from memory the protective features which can be affected by operating the level control system in the configuration given and determine which protective actions are going to occur and in what order. -. - - Basis for SRO only Level Cognitive Level QuestionType Question Source - -_____ RO Comprehension NEW Wednesday, July 14, 2010 Page 85 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION__30 - 2530 Development References udent References Provided arning Objective: )PS-ILE-DCS #17 1

References:

I) Lesson Plan OP-MC-PS-ILE-DCS_Sections 2.4.1 & 2.5.1 SYSO1 1 K3.02 Pressurizer Level Control System (PZR LCS)

Knowledge of the effect that a loss or malfunction of the PZR LCS will have on the following: (CFR: 41.7/45.6) RCS 401-9 Comments: RemarkslStatus Proposed revision for 2010 NRC Q-30. Revision appoved RFA 07/06/10. Tuesday, July 13, 2010 Page 86 of 294

Question 30 High Miss Question Proposed Replacement

References:

From Lesson Plan OP-MC-PS-ILE-DCS Section 2.4.1: When the Soft Control or the SLIMs for NV-238 is placed in Manual or the SLIMs is taken to L-MANUAL the Pressurizer Level Master Controller is swapped to Manual also by DCS. However, when the Soft Control or the SLIMs for NV-238 is returned to AUTO the operator must place the Pressurizer Level Master Controller back in AUTO. Objective #7 On the Pressurizer Level Master Controller, located on the NV CHARGING FLOW

CONTROL Graphic in DCS, the LI (Limit Increase) and LD (Limit Decrease) buttons are used to set a minimum limit LM for automatic charging flow to ensure seal injection flow to the NC Pumps is maintained. There is an LM setpoint window and also an LM bargraph displayed on the Level Master controller. The limit is set in gallons per minute. The normal setting is 35 gpm. This function is bypassed when the Pressurizer Level Master Controller or the SLIMs for NV-238 is placed in MANUAL. This function is also bypassed when the SLIMs for NV-238 is placed in L-MANUAL. This limit value is set up per OP/I (2)1A162001001A (Chemical and Volume Control System Letdown) Enc. 4.1. Objective #8 When in MANUAL, the output of the controller sets a fixed position for NV-238. Increasing the output causes NV-238 to open, while decreasing the output causes NV-238 to close. Objective #4 2.4.2 NV-238 SLIMs Station This SLIMs station is used to control the position of NV-238. In AUTO, it compares the output of the Level Master to Selected Charging Flow (which is developed using a Median Select Algorithm with three charging flow inputs) to position the valve for needed charging flow. In MANUAL or L-MANUAL, UP/DOWN push-button arrowheads are used to position the valve. When the Soft Control or the SLIMs is taken to MANUAL or the SLIMs is taken to L MANUAL the Pressurizer Master Level Controller is swapped to MANUAL also by DCS. However, when the Soft Control or the SLIMs for NV-238 is returned to AUTO the operator must place the Pressurizer Level Master Controller back in AUTO. Objective #4 2.4.3 PD Pump SLIMs Station

This station is used to control the speed of the PD Pump. The Controller will be a MANUAL only controller. The UP/DOWN arrowhead push-buttons are used to adjust speed. If the AUTO pushbutton is depressed the LED on the AUTO pushbutton will illuminate and immediately return to the MANUAL pushbutton LED illuminating. From Lesson Plan OP-MC-PS-ILE-DCS Section 2.5.1: 2.5 Control Functions 2.5.1 PZR Low Level Objective #9 In the event PZR Level decreases to 17%, valves NV1A, NV2A, NV457A, NV458A and NV35A are automatically closed. This isolates letdown to prevent further loss of inventory and minimize the possibility of uncovering the heaters. At the same time all PZR Heater groups are de-energized to protect them from overheating should they become uncovered. An Annunciator Alarm, PZR LO LEVEL HTRS OFF & LETDN SECURED, alerts the operator of the low level condition. Another feature which will isolate letdown and de-energize the pressurizer heaters is charging flow lowering to <20 gpm for> 20 seconds. The Selected Charging flow signal is developed with a Median Select algorithm with input from three (3) transmitters measuring charging flow. The low charging flow signal is maintained for 15 seconds and then clears, therefore if Pressurizer Level is >17% the Pressurizer Heaters can be placed back into service even though charging flow may not have been restored. Objective #11 Once level has increased to greater than 17% all heater groups must be manually re energized and letdown can be re-established. This is accomplished by selecting MAN on A, B, and D Heater MAN/AUTO Selector Switch. This allows closing the 600V supply breaker from their control switches on MC-5. C Heater supply breaker is closed via the switch on MC-10. There is no MAN/AUTO switch for C Heater. NOTE: If a Safety Injection has occurred, the Safety Injection signal and the sequencers must be reset in order to close the A & B heater breakers. 2.5.2 High Level Deviation Objective #9 If level should increase to greater than 5% above program level an Annunciator alarm, PZR HI LEVEL DEV CONTROL, is generated and the back-up heaters come on. This is done so that the subcooled water which has just surged into the PZR can be heated to saturation temperature. This will allow the water to flash to steam and avoid a pressure decrease as the level decreases to normal.

2.5.3 Low Level Deviation If level should decrease to less than 5% below program level an Annunciator alarm, PZR LO LEVEL DEVIATION, alerts the operator of the low level condition. 2.5.4 Hi Level Alarm If level should increase to 70% an annunciator alarm, PZR HI LEVEL, alerts the operator of the high level condition.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 31 L2 SYSO14 2.4.3 1 Rod Position Indication System (RPIS)

YS014 GENERIC ( Lnowledge of annunciator alarms, indications, or response procedures. (CFR: 41.10/45.3) Unit 1 is operating at 100% RTP. The following indications are observed on the Digital Rod Position Indication (DRPI) system:

D-4 rod indication is RED
Associated rod group background is ORANGE
IAD-2 / D10 (RPI URGENT FAILURE) is LIT Which ONE (1) of the following describes the condition of rod D-4?

A. Rod D-4 is fully inserted. B. Rod D-4 is at half accuracy. C. Rod D-4 position cannot be determined. D. Rod D-4 is greater than 231 steps withdrawn. Tuesday, July 13, 2010 Page 87 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 31 L2

  .,eneral The following indications are exhibited with a Data A and Data B failure:

a) Red position for indication for affected rods b) Red U above affected rods c) Red zero for affected rods position d) Red RB light e) Red Urgent alarm f) Orange background for affected banks g) Yellow Data Failure alarm (A and B) h) Yellow deviation alarm

1) RPI Non-Urgent Annunciator Urgent Annunciator Answer A Discussion fibORRECT: See explanation above.

PLAUSIBLE: This answer is plausible since the applicant may conclude from the indications that the rod is fully inserted and the indication is yid based on the given conditions. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: It is plausible for the applicant to conclude that the rod is at half accuracy due to a Data A OR Data B failure Answer C Discussion CORRECT: See explanation above. Answer D Discussion INCORRECT: See explanation above. LAUSIBLE: This answer is plausible if the applicant concludes that a rod which is >231 steps withdrawn gives an RPI urgent failure. DRPI is t capabile of monitoring a rod greater than 231 step withdrawn and it would be reasonable for the applicant to conclude that this condition would result in an urgent failure which would be consistent with any other condition where DRPI could not determine actual rod position. - Basis for meeting the KA iKA is matched because the applicant must understand the meaning of numerous DRPI system alarms and their impact on the operation OT the DRPI system. -__________________________________

                                                                                                                 .________

Basis for Hi Cog - iiis a higher cognitive level question. The applicant must recall what each DRPI alarm means with regards to the operation of the system. Th applicant must then analyze from the multiple alarms given in the initial conditions the overall impact on the DRPI system. Since the question qires multiple mental steps to arrive at the correct answer, this is a higher cognitive level question. Basis for SRO only Level Cognitive Level Questionyp Question Source RO Comprehension BANK CNS 2008 RO Audit Retake Exam (Q1857) Development References Student References Provided Learning Objectives: I) IC-EDA#7& 8

References:

I) Lesson Plan OP-MC-IC-EDA Section 3.2.1 SYSOI4 2.4.3 1 Rod Position Indication System (RPIS)

( S014 GENERIC nowledge of annunciator alarms, indications, or response procedures. (CFR: 41.10 /45.3) Tuesday, July 13, 2010 Page 88 of 294

________-_______ FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 31 2531 401-9 Comments: RemarkslStatus 401-9 Comments: I think the answer is obvious as written. Can you reduce the stem indications to make it less obvious? D is NP because of the 3rd bullet (Red RB is indicated for rod D-4) Replace D Resolution / Comments: Revised question and removed third bullet. Didnt really see how we could replace D with anything that was more plausible. Removing third bullet gives D plausibility. See attached file for vised copy of question. Tuesday, July 13, 2010 Page 89 of 294

Question 31

References:

From Lesson Plan OP-MC-IC-EDA Section 3.21: DRPI Urgent Alarm Objective #7,8 Refer to Drawing 7.10, D.R.P.l Display Data A + B Failure (Urgent). DRPI Urgent Alarm caused by Data A Failure and Data B Failure on a rod P-8:

The best calculated position indicated immediately below rod P-8 alpha-numaric designator would indicate a red 0.
The failure status line would indicate a red U above rod P-8 bar graph.
The rods bar graph would turn red and indicate rod height of 0.
The background color for Control Bank C would turn orange.
A red RB would be indicated on the rod bottom status line.
The system status line would indicate a yellow DATA A FAILURE, yellow DATA B FAILURE, red URGENT ALARM, and since the other rods in this bank are> 12 steps withdrawn a yellow DEVIATION> 12 STEPS.

Refer to Drawing 7.11, D.R.P.I Display Rod Deviation (Urgent) DRPI Urgent Alarm caused by an actual deviation of 12 steps on a rod P-8:

The background color for Control Bank C would turn orange.
The system status line would indicate a red URGENT ALARM and a yellow DEVIATION

        > 12 STEPS condition.

Refer to Drawing 7.12, D.R.P.l Display Gray Codes Disagree (Urgent) DRPI Urgent Alarm caused by gray codes not in agreement on rod P-8 with the result the best calculated position is 12 steps or more from other rods in the bank.

The best calculated position indicated immediately below rod P-8 alpha-numaric designator would indicate a average of Data A and Data B.
The rods bar graph would turn yellow.
The background color for Control Bank C would turn orange
The system status line would indicate a red URGENT ALARM and a yellow DEVIATION

        > 12 STEPS condition.

Note that if the gray codes not in agreement resulted in an averaged position within 12 steps of the other rods, there would be no deviation or urgent indications. The only indication would be the rod would turn yellow with a RPI Non-Urgent Annunciator. An example of this scenario is when leads for Data A and Data B are rolled, and rods are withdrawn. DRPI sees the B coil made first, knows this is a disagreement and intermittently turns the rod yellow (until the A coil is made), but the indicated position never gets 12 steps from the other rods, so no deviation and no urgent alarm.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2008 CNS RO Audit Retake Examina QUESTION 57 i7 KA KA_desc SYSO14 Knowledge of RPIS design feature(s) andior interlock(s) which provide for the following: (CFR: 41.5 / 45.7)ZRod bottom lights K4.03 Unit 1 is operating at 100% power. Given the following indications on the Digital Rod Position Indication (DRPI) system:

Associated bank background is orange
D-4 rod indication is red
Red RB is indicated for rod D-4
1AD-2, DuO RPI URGENT FAILURE is alarming Which one of the following describes the condition of rod D-4?

A. Rod D-4 is at half accuracy B. Rod D-4 at greater than 231 steps withdrawn C. Rod D-4 is fully inserted D. Rod D-4 position cannot be determined Monday, February 22, 2010 Page 113 of 151

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2008 CNS RO Audit Retake Examina QUESTION 57 0 General Discussion The following indications are exhibited with a Data A and Data B failure: a) Red position for indication for affected rods 1 b) Red U above affected rods c) Red zero for affected rods position d) Red RB light e) Red Urgent alarm f) Orange background for affected banks g) Yellow Data Failure alarm (A and B) h) Yellow deviation alarm i) RPI Non-Urgent Annunciator

 ) RPI Urgent Annunciator Answer A Discussion Plausible: The student may believe the rod is at half accuracy due to a Data A OR Data B failure Answer B Discussion rPlausible: The student may believe that rod is >231 steps withdrawn gives an RPI urgent failure.

Answer C Discussion [plausible: The student may believe the rod is fully inserted and the indication is valid based on the given conditions Answer D Discussion Correct: The indications given are for a Data A and Data B failure for rod D-4 Basis for meeting the KA LBasis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 2007 Audit Examination#2 Q53 (Bank 53) Development References Student References Provided EDA KA KA_desc SYSO 14 Knowledge of RPIS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.5 I 45.7)ERod bottom lights K4.03 401-9 Comments: RemarkslStatus Monday, February 22, 2010 Page 114 of 151

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 32 SYSO15 K2.O1 Nuclear Instrumentation System (NIS)

nowledge of bus power supplies to the following: (CFR: 41.7) JS channels, components, and interconnections Given the following conditions on Unit 1:

Unit is shutdown in MODE 6 for Refueling
While responding to a series of alarms associated with the NIs the operator notices that the Instrument Power and Control Power lights on the PR N43 drawers are DARK Which ONE (1) of the following is the cause of these indications?

A. Inverter 1 EVIA has tripped. B. The feeder breaker for panelboard 1 EKVB has tripped. C. Inverter 1 EVIC has tripped. D. The feeder breaker for panelboard I EKVD has tripped. Tuesday, July 13, 2010 Page 90 of 294

______ FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 32 2532 C

.,eneral Discussion NIS Channel 3 (PRN43) is powered from IEKVC which is fed from Static Inverter 1EVIC.

Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because Static Inverter 1EVIA supplies panelboard 1EVCA which powers MS Channel 1 (N31. N35. and N41). Answer B Discussion -__________ __ RRECT: See explanation above. SIBLE: This answer is plausible because panelboard 1EKVB_provides power to NIS Channel 2 (N32, N36, and N42). Answer C Discussion CORRECT: See explanation above. Answer 0 Discussion INCORRECT: See explanation above. PLAUSIBLE:_This answer is plausible because panelboard 1EKVD provides power to NIS Channel IV (N44). Basis for meeting the KA The KA is matched because the applicant must know the power supplies for all NIS channels to determine the correct answer. Basis for Hi Cog

-_.

Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Learning Objectives: EL-EPL #5 & 6

References:

1. Lesson Plan OP-MC-EL-EPL Section 1.2 son Plan OP-MC-IC-ENB Section 2.7 SYSO15 K2.01 Nuclear Instrumentation System (NIS)

Knowledge of bus power supplies to the following: (CFR: 41.7) NIS channels, components, and interconnections Comments: RemarkslStatus 401-9 Comments: No comment. Resolution / Comments: N/A Tuesday, July 13, 2010 Page 91 of 294

Question 32

References:

From Lesson Plan OP-MC-EL-EPL Section 1.2: 120 VAC Vital Instrumentation and Control Power System Objective # I The 120 VAC Vital Instrumentation and Control Power System consist of four vital panelboards and four inverters to each unit. The four vital panelboards will normally receive power through static inverters 1(2) EVIA, 1(2) EVIB, 1(2) EVIC and 1(2) EVID. A regulated power supply (1 KRP for Unit I and 2 KRP for Unit 2) is also provided, as an alternate power source, to allow uninterruptible manual power transfer to panelboards 1(2) EKVA, 1(2) EKVB, 1(2) EKVC, and 1(2) EKVD when an inverter is intentionally taken out-of-service. This system provides four independent channels for instrumentation and control power to both units (Unit 1 and 2). A Train loads are fed from channels A and C while the B Train loads are fed from channels B and D. Three of the four channels will ensure that the overall system functional capability is maintained, comparable to the original design standards for safe operation. However, a loss of any two of these channel sources will result in a shutdown of the respective unit. Objective # 6 The following is a listing of typical loads that are powered from the 120 VAC Distribution Centers:

NIS Channels 1 thru 4 Instrument Power
NIS Channels 1 thru 4 Control Power
SSPS Instrument Power
SSPS Control Power
FWST Channels 1 thru 4 Instrument Power
Containment Radiation Monitors Isolation Valves
Auxiliary Safeguard Cabinets Instrument Power
Post Accident Recorders
Post Accident Annunciators tO COMPONENT DESCRIPTION 1.1. 125 VDC Vital Instrumentation and Control Power System Battery Chargers The two-unit station is provided with five battery chargers, designated EVCA, EVCB, EVCC, EVCD; and a spare battery charger, designated EVCS, which can be used to replace a charger if required. These chargers, supplied by SCI (Solid state Controls Incorporated), are 500 ampere chargers with a charging capability of 500-625 amps, however, we have them current limited at 525 amperes.

From Lesson Plan OP-MC-IC-ENB Section 2.7: 2.7 Power Supplies NIS Channel I EKVA NIS Channel II EKVB NIS Channel Ill EKVC NIS Channel IV EKVD Wide Range Train A EKVA Wide Range Train B EKVD 3.0 SYSTEM OPERATION 3.1 Normal Operation 3.1.1 Operating Procedures The Excore Nuclear Instrumentation System provides the operator with neutron flux indication for all modes of operations. During each reactor startup, a healthy skepticism (N concerning the validity of power indications is warranted, particularly following a V refueling outage. Changes in plant equipment or conditions, along with a strong desire to return the plant to full operation, may influence personnel to accept less than complete explanations for discrepant indications. For example, excessive electrical generation for the nuclear power indicated (a symptom of miscalibrated nuclear instruments) has been attributed to factors such as: cold circulating water temperature, expected efficiency improvements, and changes in core design or instrumentation.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 33 SYSO16 K4.Ol Non-Nuclear Instrumentation System (NNIS)

nowledge of NNIS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) {eading of NNIS channel values outside control room Which ONE (1) of the following sets of indications can be read outside the Main Control Room on BOTH the Auxiliary Shutdown Panel (ASP) AND the Safe Shutdown Facility (SSF) Control Panel? A. SR Neutron Flux AND SIG WR Levels B. SR Neutron Flux AND Pressurizer Level C. Incore Thermocouples AND SIG WR Levels D. Incore Thermocouples AND Pressurizer Level Tuesday, July 13, 2010 Page 92 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 33 eneraI Discussion Pressurizer level and SR Neutron Flux can be read outside the Main Control Room on both the ASP and SSF Control Panels. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because SR Neutron Flux can be read on the both the ASP and the SSF. SG Wide Range level can be read on the SSF but not on the ASP. Answer B Discussion RRECT:_See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE:_This answer is plausible because S/G WR Levels and Incore Thermocouples can both be read on the SSF Control Panel. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because Incore Thermocouples can be read on the SSF Control Panel and Pressurizer level can be read on both the ASP and SSF.

                                                                                                 -                                _______

Basis for meeting the KA The KA is matched because the applicant must recall all indications (both Nuclear and Non-Nuclear indications) available at the SSF and ASP. Basis for Hi Cog Basis for SRO only cJZZZ Job Level Cognitive Level QuestionType Question Source L RO Memory NEW Development References Student References Provided Learning Objectives:

1. CP-ASP #2

2. CP-AD #8

References:

I. Lesson Plan OP-MC-CP-AD Section 2.1

2. Lesson Plan OP-MC-CP-ASP Section 2.1 SYSO 16 K4.0 1 Non-Nuclear Instrumentation System (NNIS)

Knowledge of NNIS design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) Reading of NNIS channel values outside control room 401-9 Comments: RemarkslStatus 401-9 Comments: Consider adding an additional indication to increase LOD. A,A (none good) A,B (B good only) B,B (All good) B,A (B good only Resolution / Comments: Developed a revised question with two answers per distracter. If Tuesday, July 13, 2010 Page 93 of 294

____ FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 33 revised question is used the distracter analysis will need to be vised._See attached file for revised copy of question. Tuesday, July 13, 2010 Page 94 of 294

Question 33

References:

From Lesson Plan OP-MC-CP-ASP Section 2.1: 1.0 COMPONENT DESCRIPTION 1.1. Panel Indications (Refer to Drawings 7.1, 7.2, 7.3, & 7.4) 1.1.1. Temperature indications (all temperature indication is continuous)

Reactor Coolant System Wide Range Hot Leg Temperature (0-700°F) Loop D Hot Leg
Reactor Coolant System Wide Range Cold Leg Temperature (0-700) Loop D Cold Leg
Regenerative Heat Exchanger Letdown Temperature (100-600)
A & B ND Pump Discharge Temperature (50-400°F)
A, B, C, & D ND to Cold Leg Temperatures (50-400°F) 1 .1 .2. Pressure Indications (all pressure indication is continuous)
Wide Range Reactor Coolant System Pressure (0-3000 psig)
Narrow Range Reactor Coolant System Pressure (PZR Press) (1700-2500 psig)
Letdown Pressure (0-600 psig) 1.1.3. Level Indications (continuous)
Channel 1 Pressurizer Level (0-100%)

1.1.4. Flow Indications (continuous)

Letdown Flow (0-200 gpm) 1 .1 .5. Power Indication (continuous)
SR Nuclear Flux (10-1 - I Q5 cps, separately detected, not part of the NIS) 1.2. Manual Loaders on the Panel 1.2.1. NV-459 (Variable L/D Orifice Outlet Flow Control)

It is used to throttle letdown flow rate when initiating letdown. Throttling prevents thermal shock of the letdown piping by allowing the operator to slowly initiate letdown. Per procedures, excess letdown is established if normal letdown is not in service. 1.2.2. NV-21A (NV Spray to PZR Isol) It is used to control NC System pressure if the normal spray valves are unavailable or not functioning properly (Note: the normal spray valves should be operating in Auto, and no control of them on the ASP). Its used on the ASP for NC System pressure control during cooldown. Letdown must be in service before this valve can be used. This is to ensure the AT between the Pressurizer Temperature and Spray Water is less than 320°F, which aids in preventing thermally shocking the spray nozzle. When NV-21 is being used, valves NV-13B and 16A must be closed (Normal and Alternate Charging) which allows the operator to maintain a more constant letdown and charging flow balance.

From Lesson Plan OP-MC-CP-AD Section 2.1: The pump is driven by an induction motor powered from the standby shutdown power supply. Control switches for the pumps and various isolation valves are located on the SSF Control Panel. A filter is provided downstream of the pump to collect any particulate matter larger than 5 microns that could cause damage to the reactor coolant pump seals. Filter differential pressure is indicated locally. Since the makeup pumps deliver a constant flowrate to the Reactor Coolant System, it may become necessary to remove excess water to maintain Pressurizer level 60 80%. - Solenoid operated, reactor vessel head vent valves (NC272 & 273) are powered by the Standby Shutdown system to allow discharge of water to the Pressurizer Relief Tank (PRT). Controls for these valves are located on the SSF Control Panel. A flowpath for the Standby M/U Pump is provided by opening NV842AC and NV849AC. These valves will close on a Phase A (Si) signal if they are being powered from their normal power supply (EMXA-4). Once control is swapped to the SSF and EM)(A-4 is swapped to its alternate power supply (MCC SMXG) the valves will no longer close on a Phase A (Si) signal. Pressurizer level is indicated on the SSF Control Panel. 1.2.3. Temperature Indication Five Core Exit Thermocouples can be monitored from the SSF Control Panel to monitor A multi-conductor cable that is connected on the side of the control panel must be relocated in order to view the thermocouple readings. The highest reading Core Exit Thermocouple is used to determine subcooling. Indication is also provided for the Incore reference junction temperature deviation. This temperature deviation indication is used to obtain a corrected Core Exit Thermocouple value to be used in determining subcooling. Indication is also provided for Loop A and D WR Cold Leg temperatures.

1.2.4. Pressure Control In order to prevent steam bubble formation in the reactor vessel, primary pressure must be maintained above saturation pressure at the core exit temperature. A sub-group of Back-Up Heater Group D (7O kW) is powered from the SSF electrical distribution system and can be controlled from the SSF Control Panel. The heaters are energized as necessary to maintain subcooling if pressure decreases. This ensures the steam bubble stays in the Pressurizer. The heaters have a LOCAL/REMOTE switch and a control switch. The LOCAL position bypasses all AUTO and Control Room functions. The Pressurizer Spray valves can also be controlled from the SSF Control Panel. The spray valves have open/close switches which are used to ensure that the spray valves remain closed (gives a hard closed signal). The normal position for this switch is the closed position. This switch is only functional when controlling (via EMXA-4 swap) from SSF. Reactor Coolant System wide range pressure indication is provided on the SSF Control Panel. 1.2.5. Flux Indication WR Neutron Flux Indication is provided on the SSF Control Panel. Indication is provided from 1 O CPS up to I O CPS. 1.3. Secondary System Control Steam Generator Wide Range level indication is provided on the SSF Crro.an These level indicators are calibrated for hot conditions since the design of the SSF is to mntaiWHot Standby. The TD CA Pump will auto start if 1/1 WR level transmitter indicates 72% on 2/4 S/Gs. A step in the body of AP-24 Loss of Plant Control due to Fire or Sabotage will have the operator manually start the TD CA pump prior to leaving the control room and a step in AP-24 Enc. 1 will place SA-48ABC in the open position at the SSF. Procedurally the TDCA flow will be controlled based on the availability of the controls in this order: control room, the CA pump room or locally in the dog houses. A steam supply is assured to the TD CA Pump on swap over to the SSF due to the MSIV and S/G PORV on C S/C failing closed. Feedwater is also assured to provide a heat sink due to the CA supply valves (CA 54AC and CA 66AC) from the TD CA Pump to B and A S/Cs failing as is (with a normal position of open) on swap over to the SSF. Feedwater is provided to C S/G from the TD CA Pump by verifying CA5OB (TD CA to S/G C Isol) open and securing the hand wheel clutch in the engaged position as directed by procedure. NOTE: The word disengaged in the next paragraph refers to the motor not the handwheel. Permanently installed step ladders were added in the basement of the doghouse near CA54AC and CA66AC. The motor operator clutch levers for CA38B, CA5OB, CA54AC,

and CA66AC have eyelets such that an eyebolt can be screwed into them to secure the lever in the disengaged position. The eyebolts are stored on the clutch lever plates using a short piece of small wire. Labels are attached with this wire which indicates that eyebolts are dedicated for use during certain SSF Events. A two position switch for SA-48ABC (A FWDT Steam Supply) is located on the SSF Control Panel to prevent continual cycling of the TD CA Pump. The two positions are: AUTO: SA-48ABC will open in response to an auto start signal. OPEN: Seals in SA-48ABC in the open position and bypasses the auto start signal. This switch will normally be maintained in the Auto position. It will be selected to open by Enclosure 1 of AP-24 to seal in the auto start signal to the TDCA pump. NOTE: This switch will only affect the SSF related solenoids.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 34 SYSO28 A2.O1 Hydrogen Recombiner and Purge Control System (HRPS)

alfunctions or operations on the HRPS; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of jose malfunctions or operations: (CFR: 41.5/43.5/45.3 /45.13) Hydrogen recombiner power setting, determined by using plant data book Given the following on Unit 1:

A LOCA occurred 24 hours ago
The 1A H2 Recombiner was placed in service per EP/lIN5000/Gl Enclosure 4 (Placing H2 Recombiners In Service)
Containment pressure was 5 PSIG when the Recombiner was placed in service Current Conditions are as follows:
Containment pressure is 1 .5 PSIG Based on the conditions above the recombiner Power Setting was (1) when the recombiner was placed in service and should now be set to (2)

Which ONE (1) of the following completes the statement above? REFERENCE PROVIDED A. 1. 49.8KW

2. 45.3KW B. 1. 49.8KW

2. 45.8KW C. 1. 50.3KW

2. 45.3KW D. 1. 50.3KW

2. 45.8KW Tuesday, July 13, 2010 Page 95 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 34 2534 General Discussion In the scenario given with this question the power setting for the H2 recombiner should initially be 49.8 KW based on the initial containment pressure. Based on the current containment pressure of 1.5 PSIG the power setting should be 45.8 KW. Initial Power setting - Pressure Factor, CP = 1.395 @ PSIG Reference Power 3 5.670 KW Power Setting CP x Reference Power Power Setting = 1.395 x 35.67 = 49.8 KW Current Power setting - Pressure Factor, CP = 1.285 @ 5 PSIG Reference Power = 35.670 KW Power Setting CP x Reference Power Power Setting= 1.285 x 35.67 45.8KW Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is correct. Part 2 is plausible if the applicant incorrectly reads the wrong pressure line on the graph. Incorrectly reading the graph is plausible since the major divisions are in 2 PSIG increments and the minor divisions are in 1/2 PSIG increments. nswer B Discussion ZRRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: Both parts are plausible if the applicant incorrectly reads the wrong pressure line on the graph. Incorrectly reading the graph is plausible since the major divisions are in 2 PSIG increments and the minor divisions are in 1/2 PSIG increments Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible if the applicant incorrectly reads the wrong pressure line on the graph. Incorrectly reading the graph is plausible since the major divisions are in 2 PSIG increments and the minor divisions are in 1/2 PSIG increments Part 2 is correct Basis for meeting the KA The KA is matched because the applicant is asked to determine the Power Setting for the recombiner under two different conditions. This requires the applicant to determine the Pressure Factor both conditions using the Power Correction Factor curve from the Plant Data Book and then calculate the correct Power Setting for each condition. Basis for Hi Cog This is a hi cog question because the applicant must read the Power Correction Factor graph from the Plant Data Book and use the information from the graph to calculate the correct power setting. Since this requires more than one mental step, it is a higher cognitive level question. Basis for SRO only --_______________ Job Level Cognitive Level QuestionType Question Source - RO Comprehension NEW ...ievelopment References Student References Provided Lesson Plan OP-MC-CNT-VX Pg. 27 (Rev 23) U-l Data Book Curve 1.8 EP Generic Enc G-l End. 4 OP-MC-CNT-VX Obj.7 Tuesday, July 13, 2010 Page 96 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 34 SYSO28 A2.O 1 Hydrogen Recombiner and Purge Control System (HRPS)

alfunctions or operations on the HRPS; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of nose malfunctions or operations: (CFR: 41.5 / 43.5/45.3/45.13) Hydrogen recombiner power setting, determined by using plant data book 401-9 Comments: - - RemarkslStatus 401-9 Comments: No comment. Resolution / Comments: Replaced should in stem of question with will based on General Comment from Lead Examiner. See attached copy of question for proposed revision. Tuesday, July 13, 2010 Page 97 of 294

Question 34

References:

NNLLL OBJECTIVE L L P P 0 OORSR ROO

7. Discuss the instrumentation and controls associated with the X X X X X Hydrogen Recombiners, to include:

      . Temperature readout
      . Power adjust potentiometer
      . Power out meter
      . Power out switch
      . Power available light.

8. Discuss the instrumentation associated with the Hydrogen X X X X X Analyzer Concentration Monitors.

g Evaluate plant parameters to determine any abnormal system x x x x x conditions that may exist.

10. Given a limit and/or precaution associated with an Operating X X X X X Procedure, discuss its basis and applicability.

11 Concerning the Technical Specifications related to the VX System:

      . Given the LCO title, state the LCO (including any         X X X COLR values) and applicability.
      . For any LCOs that have action required within one        x x x hour, state the action.
      . Given a set of parameter values or system conditions,     x x x determine if any Tech. Spec. LCOs is (are) not met and any action(s) required within one hour.
      . Given a set of parameter values or system conditions and the appropriate Tech Spec, determine required action(s)
      . Discus the bases for a given Tech. Spec. LCO or

Safety Limit SRO ONLY

From OP-MC-CNT-VX Pg. 27 (Rev 23) 3.2 Abnormal and Emergency Operation The control panels for the electrical Recombiners are located in the MG set rooms. The recombiner units are located in Containment such that they process a flow of Containment air containing hydrogen at a concentration typical of the lower containment compartments. This is because the Hydrogen Skimmer Fans discharge in the vicinity of the Recombiners and the Recombiners process that flow. There is no direct piping or duct connection between the Recombiners and the Hydrogen Skimmer Fans. The recombiner consists of a thermally insulated vertical metal duct with electric resistance metal-sheathed heaters provided to heat a continuous flow of Containment air (containing a low concentration of hydrogen), up to a temperature which is sufficient to cause a reaction between hydrogen and oxygen (between 1225°F and 1400°F). The recombiner is provided with an outer enclosure to keep out water coming from the Containment Spray System. The recombiner consists of an inlet preheater section, a heater-recombination section, and a mixing chamber. The warmed air passes through an orifice plate (should protect the recombiner from being overloaded from higher hydrogen concentrations up to 6.0%) and then enters the electric heater section where it is heated to approximately 1225-1400°F causing recombination to occur. Tests have verified that the recombination is not a catalytic surface effect associated with the heaters, but occurs due to the increased temperature of the process gases. Since the phenomenon is not a catalytic effect, poisoning of the unit as by fission products will not occur. The heater section consists of five assemblies of electric heaters stacked vertically. Each assembly contains individual heating elements. Operation of the unit is virtually unaffected in the event of a few individual heating elements failing to function properly. Objective #7 The recombiners are equipped with chromel-alumel thermocouples with a reference junction monitored with an RTD. Digital temperature meters are provided on the Hydrogen Recombiner Heater Temperature Monitor Panel ( refer to Drawing 7.3) located in the MG set rooms. The display is normally off but may be operated if desired by:

1) Power on

2) Unit will perform self diagnostics and Return: Command?,

3) Press AUTO key.

The unit will display sequentially the three thermocouples points (numbered 1, 2, 3) and the reference junction temperature (number 4). The value for the reference junction is not fixed and is used to perform reference junction compensation for the thermocouples inputs. The three thermocouples provide recombiner temperature indication during testing. Temperature indication is not required during a LOCA, so the thermocouples portion of the recombiners is non-safety related, and both trains are on the same panel.

Question 34

References:

NNLLL OBJECTIVE L L P P 0 OORSR ROO

7. Discuss the instrumentation and controls associated with the X X X X X Hydrogen Recombiners, to include:

      . Temperature readout
      . Power adjust potentiometer
      . Power out meter
      . Power out switch
      . Power available light.

8. Discuss the instrumentation associated with the Hydrogen X X X X X Analyzer Concentration Monitors.

. Evaluate plant parameters to determine any abnormal system x x x x x conditions that may exist.

10. Given a limit and/or precaution associated with an Operating X X X X X Procedure, discuss its basis and applicability.

11 Concerning the Technical Specifications related to the VX System:

      . Given the LCO title, state the LCO (including any         X X X COLR values) and applicability.
      . For any LCOs that have action required within one        x x x hour, state the action.
      . Given a set of parameter values or system conditions,     x x x determine if any Tech. Spec. LCOs is (are) not met and any action(s) required within one hour.
      . Given a set of parameter values or system conditions and the appropriate Tech Spec, determine required action(s)
      . Discus the bases for a given Tech. Spec. LCO or

Safety Limit SRO ONLY

From OP-MC-CNT-VX Pg. 27 (Rev 23) 3.2 Abnormal and Emergency Operation The control panels for the electrical Recombiners are located in the MG set rooms. The recombiner units are located in Containment such that they process a flow of Containment air containing hydrogen at a concentration typical of the lower containment compartments. This is because the Hydrogen Skimmer Fans discharge in the vicinity of the Recombiners and the Recombiners process that flow. There is no direct piping or duct connection between the Recombiners and the Hydrogen Skimmer Fans. The recombiner consists of a thermally insulated vertical metal duct with electric resistance metal-sheathed heaters provided to heat a continuous flow of Containment air (containing a low concentration of hydrogen), up to a temperature which is sufficient to cause a reaction between hydrogen and oxygen (between 1225°F and 1400°F). The recombiner is provided with an outer enclosure to keep out water coming from the Containment Spray System. The recombiner consists of an inlet preheater section, a heater-recombination section, and a mixing chamber. The warmed air passes through an orifice plate (should protect the recombiner from being overloaded from higher hydrogen concentrations up to 6.0%) and then enters the electric heater section where it is heated to approximately 1225-1400°F causing recombination to occur. Tests have verified that the recombination is not a catalytic surface effect associated with the heaters, but occurs due to the increased temperature of the process gases. Since the phenomenon is not a catalytic effect, poisoning of the unit as by fission products will not occur. The heater section consists of five assemblies of electric heaters stacked vertically. Each assembly contains individual heating elements. Operation of the unit is virtually unaffected in the event of a few individual heating elements failing to function properly. Objective #7 The recombiners are equipped with chromel-alumel thermocouples with a reference junction monitored with an RTD. Digital temperature meters are provided on the Hydrogen Recombiner Heater Temperature Monitor Panel (refer to Drawing 7.3) located in the MG set rooms. The display is normally off but may be operated if desired by:

1) Power on

2) Unit will perform self diagnostics and Return: Command?,

3) Press AUTO key.

The unit will display sequentially the three thermocouples points (numbered 1, 2, 3) and the reference junction temperature (number 4). The value for the reference junction is not fixed and is used to perform reference junction compensation for the thermocouples inputs. The three thermocouples provide recombiner temperature indication during testing. Temperature indication is not required during a LOCA, so the thermocouples portion of the recombiners is non-safety related, and both trains are on the same panel.

MNS GENERIC ENCLOSURES PAGE NO. EP!i 1AIS000!G-i 17 o 168 Enclosure 4 Page 1 of 8

UNIT 1 Placing H2 Recombines In service Rev. 5 Acncw.*Exp::Tr a:2rcusE REsPoTSE ior camnzi

1. Select one train of 1-12 Recombiner to be placed in service:

   . To start 1A H2 Recombiner, GO TO Step 2.

OR

To start lB H2 Recombiner, GO TO StepS.

2. Determine 1A H2 Recombiner power setting as follows:

a. Determine PRESSURE FACTOR, CP from Data Book Curve 1.8.

b. Multiply lA REFERENCE POWER listed on Data Book Curve 1.8 by PRESSURE FACTOR. OP to determine 1A Hydrogen Recombiner Pover Setting.

K = lk aEF:aEHCE rotcE; PVEEu2z nc:DP.,.:; IA r:.wr Setzzrg

c. Record IA POWER SETTING

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 35 2535 SYSO33 Al.02 Spent Fuel Pool Cooling System (SFPCS)

                        -
     \bility to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with Spent Fuel Pool Cooling System
    .)perating the controls including: (CFR: 41.5 / 45.5)

Radiation monitoring systems Given the following events and conditions associated with the Unit I SFP:

A Lo-Lo alarm is received for OAC point M1A0004 (SFP Level)
The operators read (-)2.i ft SFP level and steady on the main control board
The operating KF pump has tripped
An NEC reports a large leak in the auxiliary building but the leak has now slowed to a trickle For the event described above the leak must be associated with the KF pump (1) piping and (2) would be utilized to monitor increasing radiation levels associated with the loss of SEP level.

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

2. IEMF-42 (U-i Spent Fuel Bldg Vent)

B. 1. discharge -

2. IEMF-17 (Spent Fuel Bldg Refuel Brdg)

C. 1. suction

2. IEMF-42 (U-i Spent Fuel Bldg Vent)

D. 1. suction

2. IEMF-17 (Spent Fuel Bldg Refuel Brdg)

Tuesday, July 13, 2010 Page 98 of 294

________________ FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 35 2535 General Discussion In the scenario described in the stem of this question, the indication provided would be consistent with a SFP cooling system leak associated with the discharge piping. The design of the piping is such that the hole drilled in the discharge piping located 2 feet below the normal level (Indication of 0 feet) act as a siphon breaker. With the pump tripped, this type of leak should slow to a trickle once level goes below this value. IEMF-17 is an area monitor located on the refueling bridge and would be the most direct indication of any increase in rad levels associated with the falling SFP level. IEMF-42 uses a beta gas detector which monitors the SFP ventilation rad levels. However, IEMF-42 is designed to detect fuel failure based on the release of fission product gases. 1EMF-42 is located in the ventilation ducting in another building from the SFP and is shielded from background radiation. For the scenario_described, there would be no effect on 1EMF-42 indication. Answer A Discussion - INCORRECT: See explanation above. PLAUSIBLE: Part (1) is correct and therefore plausible. Part (2) is plausible because it does monitor radiation levels associated with the SFP building ventilation system and if the applicant misinterprets the indicated level to be low enough to cause extreme radiation level this would be a reasonable answer. - Answer B Discussion

                                                                                              .                   --_____

CORRECT: See explanation above. Answer C Discussion - INCORRECT: See explanation above. PLAUSIBLE: Part (1) is plausible if the applicant confuses the siphon breaker location to be on the suction piping verses the discharge piping. (Part (2) is plausible because it does monitor radiation levels associated with the SFP building ventilation system and if the applicant isinterprets the indicated level to be low enough to cause extreme radiation level this would be a reasonable answer. swer D Discussion CORRECT: See explanation above. PLAUSIBLE: Part (1) is plausible if the applicant confuses the siphon breaker location to be on the suction piping verses the discharge

*piping.

Part (2) is correct and therefore plausible. Basis for meeting the KA There is no direct correlation between the ability to monitor Radiation Monitor System parameters to prevent exceeding design limits associated with the Spent Fuel Pool Cooling System. However, for this particular question the applicant is asked to evaluate a given set of conditions and predict the minimum design SFP level which would be expected if leak developed on the discharge piping for the Spent Fuel Pool cooling pump. Additionally, the applicant is asked to identify which EMF could be used to verify the presence of a leak and that the leak has stopped. For example, in addition to the fact that SFP level has stopped decreasing, the Operator could use 1EMF-17 as rad monitor indication would initially increase due to lowering SFP level and then stop increasing when the leak_stops). Basis for Hi Cog This is a hi cog question because it involves a level of analysis of given situation, apply system knowledge and solve a problem. Basis for SRO only Hob Level Cognitive Level QuestionType Question Source RO Comprehension BANK Bank MNS FI-IKFNO1 eveIopment References Student References Provided

    ;son Plan OP-MC-FH-KF Page 27 (Rev 30)

OP.-MC-FH-KF Obj. 7 _. SYSO33 A 1.02 Spent Fuel Pool Cooling System (SFPCS)

Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with Spent Fuel Pool Cooling System Wednesday, July 14, 2010 Page 99 of 294

________ _______ FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 35 2535 operating the controls including: (CFR: 41.5 / 45.5) diation monitoring systems Comments: RemarksiStatus r4019 - 40 1-9 Comments: The distractor analysis said 1EMF-42 will have little to no effect. If A is marginally correct then it can be arguably correct. Therefore, 2 potentially correct answers exist. This must be re-evaluated. This Q is U until resolved due to 2 possible correct answers. Resolution / Comments: The discussion should have stated that 1EMF-42 will have no effect instead oflittle to no effect. This event would be dealt with via entry into AP-4l (Loss of Spent Fuel Pool Cooling or Level). An alarm on I EMF- 17 is one of the symptoms that prompts entry into AP-41. There is plausibility for 1EMF-42 in that an alarm on this monitor would prompt entry into AP-25 (Spent Fuel Damage). However. 1EMF-42 is a beta gas monitor and will only respond if there is damage to the fuel in the SFP. Revised the discussion and distracter analysis for A2 and C2. See attached file for proposed changes to the discussion and di stracter_analysis. Tuesday, July 13, 2010 Page 100 of 294

Question 35

References:

OP-MC-FH-KF Obj. 12 OBJECTIVES NNLLL OBJECTIVE L L P P 0 OORSR ROO I State the purpose of the Spent Fuel Pool Cooling System. X X X X 2 Draw a simplified diagram of the Spent Fuel Pool Cooling X X X X System (including all major components) per Training Drawing 7.1, Spent Fuel Pool Cooling System Simplified.

3 State the flowrates through each of the following flowpaths: X X X X

   . Spent Fuel Pool Cooling Loop
   . Spent Fuel Pool Purification Loop
   . Spent Fuel Pool Skimmer Loop 4   List the sources of makeup to the Spent Fuel Pool Cooling      X X X X System; including the source grade (i.e., borated, non-borated demineralized, and non-borated lake water).

5 Explain the conditions which would require assured makeup X X X X X from the Nuclear Service Water System, to the Spent Fuel Pool Cooling System. 6 List the power supply for the following Spent Fuel Pool X X X X Cooling System Pumps (Unit 1 and Unit 2):

   . KF Pump(s)
   . KF Skimmer Pump(s)

Describe the controls, indications, and/or alarms, associated x x x with Spent Fuel Pool Cooling System operation, located within the Control Room. 8 Describe how the KF Pump motor(s) is cooled during system X X X X operation. 9 State the cooling medium for the Spent Fuel Pool Cooling X X X X System Heat Exchanger(s). i Describe the controls, indications, and/or alarms, associated x x x x with Spent Fuel Pool Cooling System operation, located outside the Control Room.

From Lesson Plan OP-MC-FH-KF Page 27 (Rev 30) The Spent Fuel Pool stores fuel assemblies approximately 33 feet 4 inches below the fuel pool operating deck with approximately 25 feet of borated water above the top of each fuel assembly. Objective # Control Room Indication is provided for Spent Fuel Level and Temperature. (Refer to Training Drawing 7.3, Spent Fuel Pool Control Room Indication.) In each of the Spent Fuel Pools and refueling cavities there is an Aztec Level Gauge. The angle iron pointing out into the water is at elevation 771 43/4. This is the normal design level and corresponds to 0 on the gauge in the Control Room. Each step on the side edge of the gauge is two inches. (Refer to picture 7.5) 2.2 Spent Fuel Pool Cooling Pumps Objective # 7 Two Spent Fuel Pool Cooling Pumps (KF Pumps) are provided for each Unit. The controls and indications, associated with Spent Fuel Pool Cooling Pump operation, located on the Main Control Board (MC-11), consist of the following:

START / STOP Control Switch These momentary START / STOP pushbuttons allow the operator to START and STOP the pump, as desired.

During a Blackout the KF Pump(s) will initially lose power (load shed) but receive a manual start permissive when Load Group 9 is loaded onto the bus. During a Safety Injection Signal, the KF Pump(s) running prior to SI will continue to run. The KF Pump(s) not running, prior to SI, will receive a manual start permissive when Load Group 9 is loaded onto the bus. Any KF Pump(s) running or manually started, while the SI Signal is present, cannot be stopped until the SI Signal is RESET.

ON 10FF (Red / Green) Indicating Lights These ON I OFF (Red / Green) indicating lights are mounted on the START! STOP Control Switch and provide indication when the KF Pump breaker is CLOSED (ON) or OPEN (OFF).

Typical flow through the heat exchanger and purification loop is 2500 gpm combined (approximately 2200 gpm through Hx and 300 gpm through purification). Each pump is designed for 3050 gpm and limited by procedure to 2900 gpm, and each takes suction from the Spent Fuel Pool, four feet below pool level, and discharge back into the Spent Fuel Pool, six feet above the fuel assemblies. Holes drilled into the Spent Fuel Pool Discharge Header act as a vacuum breaker and limit siphon draining to two feet below normal Spent Fuel Pool level.

From Lesson Plan OP-MC-FH-KF Page 53 (Rev 30) Abnormal Operating Procedure AP/1(2)/A/5500/25, Spent Fuel Damage, is provided to identify operator actions required during a spent fuel damage event. Actions are defined for spent fuel damage inside Containment or within the Spent Fuel Pool. This procedure has only a single Case and the Symptoms are:

EMF-36, Unit Vent High Gas Radiation Alarm (Process Monitor)
EMF-38, Containment High Particulate Radiation Alarm (Process Monitor)
EMF-39, Containment High Gas Radiation Alarm (Process Monitor)
EMF-40, Containment High Iodine Radiation Alarm (Process Monitor)
EMF-42, Fuel Handling High Gas Radiation Alarm (Process Monitor)
EMF-16, Containment Refueling Bridge Alarm (Area Monitor)
EMF-17, Spent Fuel Building Bridge Alarm (Area Monitor)
Gas bubbles originating from the damaged assembly(ies).
Visual evidence of damage with potential of radioactive release(s).

Subsequent operator action(s) will first determine the damaged fuel location. The area affected (Containment or the Spent Fuel Pool) must be evacuated and isolated. Those personnel evacuated must be assembled for accountability while remote action(s) are performed to further secure the event to ON-SITE. In addition, the event must be classified and implementation of the Emergency Plan initiated, if required.

Parent Question Question 375 FHKFN0I FHKFNO1 I Pt(s) Unit 1 is operating at 100% power when the OAC registers a low spent fuel pool level alarm. Given the following events and conditions:

The operators read -2.1 ft SEP level and steady on the main control board.
The operating KE pump has tripped.
An NLO reports a large leak in the auxiliary building.
Normal SEP makeup is not available.

Which one of the following statements correctly describes the corrective action for this event? A. Find and isolate the leak on the KF discharge piping. B. Find and isolate the leak on the KF suction piping. C. Initiate assured makeup due a leak on the discharge piping. D. Initiate assured makeup due to a leak on the suction piping.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 36 SYSO35 K1.O1 Steam Generator System (S/GS)

Knowledge of the physical connections and/or cause-effect relationships between the S/GS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8) MFW/AFW systems Given the following conditions on Unit 1:

A unit shutdown is in progress
Operators have blocked the CA Auto-Start signal
At 0200 both Main Feed Pumps trip Given the following plant conditions and times:

Time Condition 0200 0205 0210 0215 0220 Tave(°F) 551 552 552 553 554 NC Press. (PSIG) 1951 1953 1958 1951 1957 NRSGA(%) 24 16 25 18 10 NRSGB(%) 26 18 22 14 9 NRSGC(%) 28 20 26 13 8 NRSGD(%) 23 15 16 19 9 Which ONE (1) of the following lists the EARLIEST time that the Turbine Driven CA pump would have automatically started? A. 0205 B. 0210 C. 0215 D. 0220 Tuesday, July 13, 2010 Page 101 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 36 L General Discussion The Turbine Driven CA Pump will auto-start when NR level on any two SGs decreases to less than 17%. For the conditions given, the Turbine Driven CA Pump will auto-start at 0205. Answer A Discussion CORRECT: See explanation above. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant concludes that NC system pressure must increase above the P-l 1 setpoint to automatically unblock the CA Auto-Start Defeat AND also believes that only one SG less than 17% is required to generate a TD CA pump auto-start. However, the Auto-Start Defeat only applies to the MD CA pumps, not the TD CA pump. ______________________ Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant concludes that NC system pressure must increase above the P-li setpoint to automatically unblock the CA Auto-Start Defeat since two of the SG NR levels are less than the 17% level required for a TD CA pump auto-start. However, the Auto-Start Defeat only applies to the MD CA pumps, not the TD CA pump. Answer 0 Discussion iORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant concludes that NC system pressure must increase above the P-il setpoint to automatically unblock the CA Auto-Start Defeat AND that all four SG NR levels must be less than 17% to generate a TD CA pump auto-start signal. However, the Auto-Start Defeat only applies to the MD CA pumps, not the TD CA pump. -__________ Basis for meeting the KA The KA is matched because the applicant must understand the cause-effect relationship between SG level and the auto-start signals generated for e AFW (CA) system. 4asis for Hi Cog This is a higher cognitive level question because the applicant must associate multiple pieces of information to arrive at the correct answer. First, the applicant must recall from memory the coincidence and setpoint for the TD CA pump start and the effect of the CA Auto-Start Defeat on CA pump operation (MD CA pumps only). Then, the applicant must compare the information given in the table to the setpoint and coincidence recalled from memory to determine the correct answer. Since this question requires more than one mental step to arrive at the correct answer, it is a higher cognitive level question. Basis for SRO only Job Level Cognitive Level Question Source RO Comprehension BANK IvINS Exam Bank Question #CFCANO1O

I Development References Student References Provided Lesson Plan OP-MC-CF-CA Section 2.2 Learning Objective OP-MC-CF-CA #4 SYSO35 K1.01 Steam Generator System (S/GS)

Knowledge of the physical connections and/or cause-effect relationships between the S/GS and the following systems: (CFR: 41.2 to 41.9 / 45.7 to 45.8) MFW/AFW systems _J4O19 Comments: Remarks!Status ( 401-9 Comments: No comment. Resolution / Coments: Tuesday, July 13, 2010 Page 102 of 294

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2010 MNS SRO NRC Examination QUESTION 36

Tuesday, July 13,2010 Page 103 of 294

Question 36

References:

From Lesson Plan OP-MC-CF-CA Section 2.2: Objective #4 Refer to Figure 7.12. The auto-start signals for the CA Turbine Driven pump (which open SA-48ABC and SA-49AB) are:

2/4 detectors low-low level in any two SGs (17%)
Blackout (>8 seconds)
Ill detector from SSF SG Wide Range Low-Low Level on 2/4 SGs (72%) (only opens SA-48ABC)

NOTE: If a Blackout occurs first followed by a Safety Injection, the Sequencer will reset the start signal to the Turbine Driven CA Pump. If the Turbine Driven CA Pump is running at the time of the Safety Injection, it will continue to run. If the Safety Injection occurs first or coincident with the Blackout, the Safety Injection will BLOCK the Turbine Driven CA Pump start because the sequencer selects the Priority Mode. (This does not affect the Low-Low SG Level auto start signal or the SSF Low-Low Level start signal.) NOTE: The turbine driven pump will also start on loss of VI to the actuator or loss of power to the solenoid valves, due to the fail-open design of the valves (not considered an ESF actuation.) Operation of the following breakers may result in a Turbine Driven CA pump auto-start and a BB and NM valve auto-closure:

1 EVDA-12A, T/D Aux Feedwater Pump Train A Auto Start and Reset Controls
1 EVDD-6A, T/D Aux FWPT Train B Auto Start and Reset Controls
2EVDA-1 1A, T/D Aux Feedwater Pump Train A Auto Start and Reset Controls
2EVDD-17A, Turb Driven Aux Sol Valve 2SASVO49 Opening these breakers will result in a Turbine Driven CA Pump auto-start signal due to 1 (2)SA-48 and 1 (2)SA-49 failing open. If an auto-start signal is generated an auto closure of the BB and NM valves will occur.

The bearing oil of the TD CA Pump is cooled utilizing a small heat exchanger at the pump. The cooling medium is the fluid moving through the pump (CA system water).

Parent Question: CFCANO1 0 1 Pt During a plant shutdown on Unit 1, the operators have blocked the CA auto-start signal by depressing the auto-start defeat switch. A subsequent loss of both main feedwater pumps occurred at 0200. Given the following plant conditions at the times listed: Time Condition 0200 0205 0210 0215 0220

1) Tave(°F) 551 552 552 553 554

2) NCS pressure (psig) 1951 1953 1958 1951 1957

3) NRSGA(%) 24 16 25 18 10

4) NRSGB(%) 26 18 22 14 9

5) NRSGC(%) 28 20 26 13 8

6) NRSGD(%) 23 15 16 19 9 What time would the Turbine Driven CA Pump start automatically?

A. 0205 B. 0210 C. 0215 D. 0220 Answer 938 A Objective 4

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 37 2537 SYSO45 K5.23 Main Turbine Generator (MT/G) System

( nowledge of the operational implications of the following concepts as the apply to the MT/B System: (CFR: 41.5 /45.7) elationship between rod control and RCS boron concentration during T/G load increases Given the following conditions on Unit 1:

Reactor Power is currently being increased from 55% to 90% RTP at 3%/hr following a Refueling Outage

1. How is the withdrawal of control rods affected?

2. What changes (if any) to NCS boron concentration will be required?

REFERENCE PROVIDED A. 1. NOT restricted

2. Dilution is required.

B. 1. NOT restricted

2. Dilution is NOT required.

C. 1. Restricted

2. Dilution is required.

D. 1. Restricted

2. Dilution is NOT required.

Tuesday, July 13, 2010 Page 104 of 294

C FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 37 2537 eneral Discussion With conditions given, the plant is above the conditioned power level therefore above 40% RTP, rod withdrawal is restricted to less than 3 steps per hour per the rod maneuvering limit guidance in the U-I Data book. This restriction on Rod movements would result in additional dilutions [qiired to compensate for the negative reactivity associated with power defect during the power escalation. - - Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible if the applicant does not recall the effect of unconditioned fuel on rod movement. The applicant may conclude based on plant conditions that there is no restriction on control rod movement under the conditions given. Part 2 of the question is correct and therefore plausible. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible if the applicant does not recall the effect of unconditioned fuel on rod movement. The applicant may conclude based on plant conditions that there is no restriction on control rod movement under the conditions given. Part 2 is plausible if the applicant confuses the effect of Xenon in the scenario described in the stem. On a power escalation after a runback, 2 j dirning out and adding positive reactivity. Answer C Discussion CORRECT: See explanation above. Answer 0 Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is correct and therefore plausible. rt 2 is plausible if the applicant confuses the effect of Xenon in the scenario described in the stem. On a power escalation after a runback,

 .enon would burning out and adding positive reactivity.

Basis for meeting the KA This KJA is matched because the question is relating the effect ofaT/G load increase during an initial power escalation with unconditioned fuel. The applicant must evaluate how this would affect the relationship between Rod control and RCS boron concentration due to the limitations imposed on rod movement. Basis for Hi Cog This is a hi cog question because it involves a level of analysis of given situation, apply system knowledge and solve a problem of what both would be the effect and how the conditions given in the stem would affect operation. It also requires more than one mental step to arrive at the correct answer and is therefore a higher cognitive level question. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Lesson Plan OP-MC-CTI-I-CP (Rev 11) Pages 135, 171, 173 &175 Data Book Sect. 1.3 Enc. 4.3 OP-MC-CTH-CP Obj: 29

                                                                                                *                  ..-*__

SYSO45 K5.23 Main Turbine Generator (MT/G) System

Knowledge of the operational implications of the following concepts as the apply to the MT/B System: (CFR: 41.5 / 45.7) elationship between rod control and RCS boron concentration during T/G load increases

.J1-9 Comments:                                                                    RemarkslStatus 40 1-9 Comments:

Of the 4 bullets: I think you can delete all but the second bullet. Tuesday, July 13, 2010 Page 105 of 294

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2010 MNS SRO NRC Examination QUESTION 37 2537 In Stem 1. add at 3% per hr In Cl and Dl state Control rod withdrawal is restricted. In Al and Bi cap the word NOT Change B2 and D2 to No dilution will be required because Xenon burnout will compensate for the power defect Resolution I Comments: Deleted last two bullets. You need the first two bullets as a minimum. Made the rest of changes as recommended by Lead Examiner. See attached file for proposed revision to question. Tuesday, July 13, 2010 Page 106 of 294

Question 37

References:

-;_ Q OORSR ROO 26 Given a set of plant parameters and/or system conditions, X X X X associated with the recovery of a misaligned I dropped rod, determine the appropriate recovery limits. CTHCPO26 27 Given a set of plant parameters or system conditions, X X X X associated with the recovery of a misaligned I dropped rod, discuss the basis for the appropriate recovery limits. CTHCPO27 28 Discuss the basis for the Fuel Maneuvering Limits. X X X X CThCPO28 29 Given the Fuel Maneuvering Limits, evaluate a given se X X X X plant conditions and determine the allowable loading / ro withdrawal rates. CTHCPO29 30 Concerning the Technical Specifications related to Control Bank Insertion Limits, AFD, QPTR, and RCS Pressure, Temperature, and Flow DNB Limits:

   . Given the LCO title, state the LCO (including any COLR values) and applicability.
   . State the REQUIRED ACTION(s) and COMPLETION                      X X X TIME for action(s) with completion times of one hour or less.
   . Given a set of parameter values or system conditions,            X X X determine if any Technical Specification LCO(s) is (are) not met and any action(s) required within one hour.
   . Given a set of plant parameters or system conditions and         X X X the appropriate Technical Specification(s), determine the REQUIRED ACTION(s) and COMPLETION TIME(s).
   . Discuss the bases for a given Technical Specification            X X X LCO.

CTHCPO3O

From Lesson Plan OP-MC-CTH-CP Pg. 135 (Rev 11) POWER LEVEL Increasing reactor power (steam demand) results in two changes, one direct and one indirect, which affect power distribution: Redistribution (Direct Effect) Control Rod Movement (Indirect Effect) The first effect is the result of the variation in core AT with Reactor Power Level. As power is increased with turbine load, the core AT will rise from almost OoF at zero power to 58oF at full power. As a result the moderator in the upper portions of the core becomes progressively warmer and less dense relative to the bottom. The increasing density difference will force power toward the bottom of the core as evidenced by AFD becoming more negative The strength of this

redistribution effect is dependent on the value of the moderator temperature coefficient (MTC). At BOC when the MTC is small and negative, the redistribution effect is small. As the MTC becomes more negative with Burnup, the redistribution effect will become more pronounced. The second (indirect effect) is caused by the movement of control rods necessary to compensate, in part, for the power defect. As power is raised positive reactivity must be added in order to compensate for the negative reactivity associated with the power defect. Any rod withdrawal will tend to allow more power to be produced in the upper portions of the core, resulting in a tendency for AFD to become more positive. In practice the control rods are moved as necessary to offset the redistribution effect thereby maintaining a relatively constant axial power distribution. This is accomplished by coordinating reactor coolant boron concentration with rod position as necessary to maintain AFD on Target during the power escalation.

From Lesson Plan OP-MC-CTH-CP Pg. 171,173 & 175 (Rev 11) 3.3 Fuel Maneuvering Limits Objective # 28 The Fuel Maneuvering Limits apply to power increases ONLY. These maneuvering limits are tied to REACTOR POWER not Turbine or Generator Power. These limits are based on limiting or preventing PCI (Pellet-Clad Interaction). The primary concern is centered around previously used fuel and not new (fresh) fuel. Handling burned fuel, coupled with the fact that burned fuel has experienced fuel pellet cracking, can result in the movement of small pellet fragments. A gradual controlled increase in power will allow pellet and cladding expansion to somewhat equalize, as the fuel and cladding heat up. Objective # 29

Fuel Maneuvering Limits POWER RAMP RESTRICTIONS Recommended ramp 3% I hour, CPL 100% FP No restrictions < 40% RTP Recommended ramp =3% / hour. 40% 100% FP Max Ramp = 3%

step change, 4% in 1 hour, 7% in 2 hours, and 10% in 3 hours. OPERATING CATEGORY 1: 100% F.P. for 72 cumulative hours out of any CPL = Highest Power Level sustained fo 7 day operating period at NO at least 72 cumulative hours (consecutive or non-consecutive) out of any 7 da operating period at power. YES OPERATING CATEGORY 2: CPL = Highest Power Level sustained for at least 72 cumulative hours (consecutive or non-consecutive) out of the preceding 30 day operating period at power Recommended ramp = 3% / hour, CPL 100%

FP Max Ramp = 3% step change, 4% in 1 hour, 7% in 2 hours, wer40%F.P. YES ercP YES and 10% in 3 hours. NO CPL 100% FP

Recommended ramp = 3% / hour, CPL 100% FP Max Ramp = 3% step change, 4% in 1 hour, 7% in 2 hours, and 10% in 3 hours.

alStartupfollowing Withdraw CRs as far as satoStartup ueling Shutdown or followi... YES practical and keep CR mot itt CF RTED NO Cold Shutdown where to a minimum limiting to 3 at Low Power Levels assemblies handled. steps I hour above 50% F. YES

                                                   +
                                                                     /TRlCTlON>)

NO drawalupto50%F.P. YES Withdrawal To MaxS Of 3 steps / hour while maintaining 3% I hour Max. Power Ramp. NO rmine Condition A. A is Max. withdrawal position and corresponding power level occurring di Startup. Restrcit withdrawal to Max Rate of 3 steps / hour while maintaining 3% 1 hour Max. Ramp. For withdrawal beyond Condition A restrict withdrawal to Max Rate of 3 steps / hour and 3% I hour Max. Ramp mine Condition A. A is Max. withdrawal \TRIOTlONSon ESSARY YES position and corresponding withdrawal to Conditon A, level occurring durin, 1 Power Level, and C.R. Position. NO For withdrawal beyond inue Normal Condition A. Restrct with drawal PERATION to a Max. Rate of 3 Steps / hour and 3% I hour Max. Ramp.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 38 r__2538 SYSO7 1 K4.06 Waste Gas Disposal System (WGDS)

nowledge of design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) ( ,ampling and monitoring of waste gas release tanks Given the following plant conditions:

Waste Gas Decay Tank A is aligned for planned release
Waste Gas Decay Tank E is also mistakenly aligned for release while in service
EMF-50 (L) Waste Gas Discharge is not detecting release activity Which ONE (1) of the following would be the result if the release exceeds expected activity levels?

A. The release is monitored by 2EMF-36(L) (Unit 2 Unit Vent Gas). However, no automatic termination will occur. B. The release is monitored by 2EMF-36(L) which will automatically terminate the release if a Trip 2 alarm is reached. C. The release is monitored by 1 EMF-36(L) (Unit 1 Unit Vent Gas). However, no automatic termination will occur. D. The release is monitored by 1 EMF-36(L) which will automatically terminate the release if a Trip 2 alarm is reached. Tuesday, July 13, 2010 Page 107 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 38 2538

.,eneral Discussion The conditions given, a misalignment has resulted in EMF-50 not being aligned to properly monitor the WGDT release. The WG release is monitored by two EMFs, the primary is EMF-50 and the secondary is the U-i Unit Vent gaseous monitor 1EMF-36L. Activity detected resulting in a Trip 2 on either one of these monitors will result in a termination of the release due to the resulting auto closure of WG- 160. The

[lease will still be monitored and auto termination is still functional. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: First part is plausible if the applicant confuses IEMF-36L and 2EMF-36L. Second part is plausible if the applicant does not recall that the release can be terminated by the Waste Discharge monitor (EMF-50) or the Unit yent Monitor (1EMF-36L). Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: First part is plausible if the applicant confuses 1EMF-36L and 2EMF-36L. Second part is plausible because the Unit Vent Monitor will terminate the release on a Trip 2 alarm. However, it is 1EMF-36L instead of 2EMF-36L. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: First part is plausible if the applicant confuses 1EMF-36L and 2EMF-36L. Second part is plausible if the applicant does not recall that the release can be terminated by the Waste Discharge monitor (EMF-50) or the Unit Vent Monitor (1EMF-36L). nswer D Discussion ORRECT: See explanation above. Basis for meeting the KA This K/A is matched because the waste gas decay tank is being released and the applicant is being asked about both the design features (U-i Vent release path) and interlocks (Action for EMF Trip 2) with regard to the monitoring capability associated with the release and the tank. Basis for Hi Cog Basis for SRO only QuestionType Question Source RO Memory BANK M14S Q WEWGNO3 Development References Student References Provided Lesson Plan OP-MC-WE-WG Page 29 (Rev 12) OP-MC-WE-WG Obj. 5 SYSO7 1 K4.06 Waste Gas Disposal System (WGDS)

Knowledge of design feature(s) and/or interlock(s) which provide for the following: (CFR: 41.7) Sampling and monitoring of waste gas release tanks 401-9 Comments: RemarksIStatus 401-9 Comments: Distrsactors A and B are NP because there is no case where an isolation will not occur without a malfunction. Replace A and B. This Q is U because of 2 NP distractors. Tuesday, July 13, 2010 Page 108 of 294

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2010 MNS SRO NRC Examination QUESTION 38 Resolution / Comments: Believe there is plausibility for A and B. Will discuss. Have developed a proposed replacement question if this question is still unacceptable. See attached file for proposed replacement question. Did not use revised question. Revised distracter A in original question. Rearranged distacters. New correct answer is D. Need yorkondistracteranaIysis. Tuesday, July 13, 2010 Page 109 of 294

Question 38

References:

OP-MC-WE-WG Obj. 5 OBJECTIVES N N L L L OBJECTIVE , , R 00 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. WEWGOD3 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 WEWGDO5 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.

SROonIy

WEWGOQ7

From Lesson Plan OP-MC-WE-WG Page 29 (Rev 12) Objective #5 IWGI6O, Waste Gas Discharge Flow Controller trips closed when:

Unit vent gas IEMF36(L) (Unit I only) trip two setpoint is reached.
Waste gas OEMF5O(L) trip two setpoint is reached.

Waste Gas Radiation Monitor (2 channels) and Plant Vent Radiation Monitor are indicated on the Waste Gas Processing Panel. Alarms The following annunciators alarm on the Waste Gas Panel.

Gas Tank Pressure High (One for each of 8 tanks).
Waste Gas Moisture Separator Level High-Low (2).
Waste Gas Monitor Radiation High.
Plant Vent Monitor Radiation High.
Waste Gas Moisture Separator High Pressure (2).
Waste Gas Moisture Separator Low Pressure (2).
Waste Gas Compressor Suction Pressure Low.
Recombiner No. I Alarm.
Recombiner No. 2 Alarm.
2 Header Supply Pressure Low.

N

Primary Makeup Water to Gas Decay Tanks High Volume.
ll2 Header Supply Pressure Low.
2 Recombiner HX KC Outlet Flow Low (2).

H

Waste Gas Compressor HX KC Flow Low (2).

Any of these alarms will actuate the Waste Gas Panel Trouble Annunciator in the Control Room 0.1. Abnormal and Emergency Operation None

Parent Question WEWGNO3 Question 38 WEWGNO3 WEWGNO3 1 Pt. Given the following conditions:

Waste Gas Decay Tank A is aligned for planned release
Waste Gas Decay Tank E is also mistakenly aligned for release while in service
EMF-50 (L) Waste Gas Discharge is not detecting release activity Which one of the following would be the result of the release if the tanks exceeded Trip 2 expected levels?

A. Release will continue as an unmonitored release B. 1 EMF-36(L) (Unit 1 Unit Vent Gas) Trip 2 will secure the release C. 2EMF-36(L) (Unit 2 Unit Vent Gas) Trip 2 will secure the release D. Release is monitored, manual termination required Answer 38 B FH-KF, section 3.2

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 39 EPEOO7 EK3.01 Reactor Trip

( owledge of the reasons for the following as the apply to a reactor trip: (CFR 41.5 /41 10/45.6/45.13)

  - tctions contained in EOP for reactor trip Given the following conditions on Unit 1:

A Reactor Trip and Safety Injection have occurred due to a Small-Break LOCA
The crew has entered E-O (Reactor Trip or Safety Injection) and has reached Step 7:

Check ESF Monitor Light Panel on energized train(s) This check is performed to prevent (1) AND to (2) Which ONE (1) of the following completes the statement above? A. 1. water from entering the steam lines due to uncontrolled CA flow

2. ensure Containment release paths are isolated B. 1. excessive NC system coold own due to uncontrolled CF flow

2. ensure Containment release paths are isolated C. 1. water from entering the steam lines due to uncontrolled CA flow

2. ensure automatic actuation of Containment Spray and Containment Isolation Phase B if containment pressure exceeded 3 PSIG D. 1 excessive NC system cooldown due to uncontrolled CF flow

2. ensure automatic actuation of Containment Spray and Containment Isolation Phase B if containment pressure exceeded 3 PSIG Tuesday, July 13, 2010 Page 110 of 294

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2010 MNS SRO NRC Examination QUESTION 39 L 2539

...eneral Discussion From E-O Background Document:

STEP 7 Check ESF Monitor Light Panel on energized train(s): PURPOSE:

1. To ensure feedwater isolation has occurred.

2. To ensure non-essential containment penetrations (including ventilation penetrations) are isolated.

3. To ensure S/I pumps are running.

4. To ensure the S/I valves are properly aligned for inventory makeup.

BASIS: The ESF monitor light panel provides a quick and convenient place for the operator to check the valve positions and pump status. The CF system is isolated on a CF Isolation signal to prevent uncontrolled filling of any steam generator and the associated excessive NC cooldown that could aggravate the transient, especially if it were a steamline break. The non-essential containment penetrations are isolated to prevent potential release of radioactive materials from containment. S/I provides makeup inventory to the NC for cooling of the core during accident conditions. Since S/I is actuated, all S/I pumps have a start signal and the operator should ensure they are running. Although S/I flow is checked in subsequent steps, it is important to ensure all energized trains are properly aligned such that if one train were lost, the other train would still be available. NOTE: While the valve alignments are being checked in accordance with the Enclosures, the progress through E-O should be continued. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part I is plausible because the Feedwater Isolation is based on uncontrolled filling of the SGs. However, the basis is to prevent an uncontrolled cooldown. Other steps in E-O will provide gaining control of CA flow to prevent overfilling the SGs. Part 2 is correct and therefore plausible. Answer B Discussion CORRECT: See explanation above. iswer C Discussion iJCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because the Feedwater Isolation is based on uncontrolled filling of the SGs. However, the basis is to prevent an uncontrolled cooldown. Other steps in E-O will provide gaining control of CA flow to prevent overfilling the SGs. Part 2 is plausible because the requirement to Check ESF Monitor Light Panel on energized train(s) for containment spray and phase B is required in Step 13 of E-O once containment pressure is verified to be> 3 psig. The applicant may misinterpret this requirement to be included in the other monitor light checked contained in Step 7. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is correct and therefore plausible. Part 2 is plausible because the requirement to Check ESF Monitor Light Panel on energized train(s) for containment spray and phase B is required in Step 13 of E-O once containment pressure is verified to be > 3 psig. The applicant may misinterpret this requirement to be included in the other monitor light checked contained in Step 7. Basis for meeting the KA The KA is matched because the applicant must know the reasons for checking the ESF Monitor Light Panel during the performance of E-O (Reactor Trip or Safety Injection). Basis for Hi Cog This is a higher cognitive level question because of the high level of analysis required to arrive at the correct answer. Basis for SRO only Job Level RO { cognitive Level Comprehension QuestionType NEW

                                                                       -                             Question Source       -._________________

Tuesday, July 13, 2010 Page 111 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 39 2539 Development References Student References Provided son Plan Objective: EP-E0 #8 eferences:

1. Background Document for E-0, Step 7 EPEOO7 EK3 .01 Reactor Trip

Knowledge of the reasons for the following as the apply to a reactor trip: (CFR 41 5 /41.10 I 45.6 / 45.13) Actions contained in EOP for reactor trip 401-9 Comments: Remarks!Status 01-9 Comments: [believe this Q should be hi cog due to the fact that it involves a igher level of analysis. Resolution / Comments: Agree. Changed to big cog and provided justification. Tuesday, July 13, 2010 Page 112 of 294

Question 39

References:

From Lesson Plan OP-MC-EP-EO: STEP 7 Check ESF Monitor Light Panel on energized train(s): PURPOSE:

1. To ensure feedwater isolation has occurred.

2. To ensure non-essential containment penetrations (including ventilation penetrations) are isolated.

3. To ensure S/I pumps are running.

4. To ensure the S/I valves are properly aligned for inventory makeup.

BASIS: The ESF monitor light panel provides a quick and convenient place for the operator to check the valve positions and pump status. The CF system is isolated on a CF Isolation signal to prevent uncontrolled filling of any steam generator and the associated excessive NC cooldown that could aggravate the transient, especially if it were a steamline break. The non-essential containment penetrations are isolated to prevent potential release of radioactive materials from containment. S/I provides makeup inventory to the NC for cooling of the core during accident conditions. Since S/I is actuated, all S/I pumps have a start signal and the operator should ensure they are running. Although S/I flow is checked in subsequent steps, it is important to ensure all energized trains are properly aligned such that if one train were lost, the other train would still be available. NOTE: While the valve alignments are being checked in accordance with the Enclosures, the progress through E-O should be continued.

From Lesson Plan OP-MC-EP-EO: STEP 13 Check Containment Pressure HAS REMAINED LESS THAN 3

P51G. PURPOSE: To ensure automatic actuation of Containment Spray and Containment Isolation Phase B if containment pressure exceeded 3 PSIG. BASIS: If containment pressure exceeds 3 PSIG, containment spray is automatically initiated to mitigate the containment pressure transient. Containment Isolation Phase B valves are closed to isolate additional potential release paths from containment. A Main Steam Isolation should also occur. The RNO has the operator record the approximate time of reactor trip. The time of trip is required to be known so that operators will know when to align ND aux spray in a subsequent procedure (ES-I .3, Transfer to Cold Leg Recirc). An operator is dispatched to remove white tags and close the breakers for lNl-173A and 1NI-178B. This is done to ensure ND Aux spray is available to augment the NS spray, if and when conditions are warranted. The 50 minute time critical action to establish ND Aux spray includes restoring power to 1NI-173A and 1NI-178B. Since component cooling to the NC pump seals and motors is isolated on a Phase B signal, the NC pumps are tripped to preclude overheating of the seals and motors. However, the NC pump seal flow is maintained to assure the integrity of the NC pressure boundary. Analysis assumes that the NCPs will be secured within 10 minutes of the Phase B signal. The RV pumps are stopped because the suction flow path is isolated on the Phase B signal. The H2 Igniters are energized so that H 2 will be burned in small quantities and not be allowed to accumulate.

From E-O Step 7 MNS REACTOR TRIP OR SAFETY INJECTION PAGE NC). EPI1!AJ5000!E-O 5 of 37 UNET 1 Rev. 29

7. Check ESF Monitor Light Panel on energized train(s):

a. Groups 1,2,5 DARK.

                          -                         a. Align valves.

lx Group 3 LIT.

                     -                              b. Open valves.

c OAC IN SERVICE.

c. Perform the following:

1) Ensure bh trains Phase A Isolation are initiated.

NOTE OAC driven summary lights in Group 4 will not work. Only components wIth individual windo?.s need to be checked in next step.

2) Ensure S/I and Phase A components with individ 1131 windows inGroLip4 are lit by starting or aligning equipment as required.

3) Ensure Sf1 and Phase A valves aligned Pf the folloing, MiIe continuing in this EP:

EP/l/AJS000IG-1 (Generic Enclosures), Enclosure 1 D iS/I Valve Checklist)

                                                        . ER:IINS000/G_I (Generic Enclosures), EnclosLire ii (Phase A Valve Checklist).

4) QIQStep8.

c:zc::zccirzcczz:::zz:

7. (Continued)

d. Group 4, Rows A through F LIT AS

                                     -    d. Perfomi the following:

REQUIRED

1) Ensure both trains Phase A Isolation are initiated

2) Align or start S/I and Phase A components with individual windows in GroLip 4 as required

3) cQ IQ Step 7.f.

e. GO TO Step &

f. Check LOCA Sequencer Actuated f. Perfomi the following on energized StatLis light (181-14) on energized trains). while continuing in this EP:

train(s) LIT.

1) Ensure S/I valves aligned PER EPI I/N5000/G-1 (Generic Enclosures),

ij, Enclosure 10 (Sf1

From E-O Step 13: MNS REACTOR TRIP OR SAFETY INJECTION PAGE NO. ERl .PJ5OOOsE-O 9 of 37 UNIT 1 Rev. 29

13. Check Containment Pressure HAS

                                     -               Perform the following:

REMAINED LESS TI-lAN 3 P51G. NOTE The time of reactor trip may be used in subsequent procedures to detemine tien ND aux containment spray should be aligned.

a. Record approximate time of reactor trip.

b. Check Monitor Light Group 4, Row G.

lit.

c. jf any Row G window is dark on energized trains), THEN perform the following:

1) Initiate Phase B and Containment Spray signal.

2 IF Ro.v G sindow is still dark, THTh perform the folIoring:

                                                             . Check OAC Monitor Light Program (MONL) for Phase B, and align valves.

IF OAC is out of service. ThEN ensure Phase B valves closed PER ER(i/A/S0001G-1 (Generic Enclosures). Enclosure 12 (Phase S Valve ChecklistL while continuing in this EP.

d. Stop all NC pumps while niaintaininq

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 40 25 APEOO8 AKI.Ol Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)

                   -
.nowledge of the operational implications of the following concepts as they apply to a Pressurizer Vapor Space Accident: (CFR 41.8 /41.10/

5.3) Thermodynamics and flow characteristics of open or leaking valves Given the following conditions on Unit 1:

The unit is in MODE 3 at full temperature and pressure
The crew has entered APIIIAI5500IO1I (Pressurizer Pressure Anomalies) due to Pressurizer pressure decreasing very slowly
Pressurizer pressure is 2150 PSIG
PRT pressure is 2 PSIG Given the above conditions, determine which ONE (1) of the following would indicate a leaking PORV and the state of the fluid in the PORV discharge?

REFERENCE PROVIDED PORV Discharge Temperature State of the Effluent A. 240-280°F Saturated Vapor B. 200-240°F Saturated Vapor C. 240-280°F Wet Vapor D. 200-240°F Wet Vapor Tuesday, July 13, 2010 Page 113 of 294

_____ _______________ _______ ______ _______ ______ FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 40 General_Discussion -______ Required Reference is steam table. This question is associated with TMI. Per the TMI lesson plan: It was clear from the operators understanding of the PZR PORV discharge temperature and the indications of saturation/superheated fluid in the hot leg, that operator knowledge of thermodynamics needed to be drastically improved. At 0520 the operators obtain a printout of PZR Safety and PORV discharge temperatures showing 232°F and 283°F respectively, but the operators still believe the PORV to be closed. For some time the PORV had been leaking prior to this day. The PORV leakage had been accepted as a normal part of operation (i.e. workaround). The temperature on the discharge of the PZR PORV had indicated what would be seen for PORV open or leaking since the PORV had started leaking. The operators believed the discharge temperature would increase to PZR temperature if the PORV actually opened. A Pressurizer pressure of2 150 psig (2165 psia) corresponds to a Saturated Vapor Enthalpy of 1125 BTU/lbm. This Enthalpy undergoing a throttling process discharging to a PRT at a pressure of 2 psig (17 psia) Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE. This answer is plausible as this temperature would be obtained if the applicant follows the entropy line from the PRT press to the saturation curve. Answer B Discussion 1iCORRECT: See explanation above. PLAUSIBLE. This answer is plausible as this temperature would be obtained if the applicant follows the entropy line from the PRT pressure to the saturation curve. Answer C Discussion INCORRECT: See explanation above. AUSIBLE. This answer is plausible if the applicant follows the_entropy line from the PRT press to the saturation curve. Answer D Discussion -- iECT: See explanation above. Basis for meeting the KA - This KA is matched since the applicant must know how to use the Mollier diagram to determine the thermodynamic characteristics of the fluid entering the PRT during a Pressurizer Vapor Space Accident (i.e. leaking PORV). Basis for Hi Cog iiis is an analysis level because the applicant must evaluate the given conditions using the Mollier diagram to determine the correct temperai [nd state of the fluid. Basis for SRO only___________ Job Level Cognitive Level QuestionType Question Source Comprehension L_ BANK - 2009 NRC Q40 (Bank 2240) velopment References n!erences Provided Mollier Diagram Lesson Plan BNT-THO3R3 Steam Steam Tables Properties APEOO8 AK1.01 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)

Knowledge of the operational implications of the following concepts as they apply to a Pressurizer Vapor Space Accident: (CFR 41.8 / 41.10 / 45.3) Thermodynamics and flow characteristics of open or leaking valves

  .i1-9Cornments:                                                              ksIStatus

___ 40 1-9 Comments: No comment. Tuesday, July 13, 2010 Page 114 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination OUESTION 40 Tuesday, July 13, 2010 Page 115 of 294

opy 1.6 17 S ILWIIIII.[ Question 40 References 1540 I 1300 1500 26,

                                ,                                                                               1470 1440 1430 1420 I 1310 3*0 br                                                                                     *210 210 0210 1260 5250 1240 1210 III Ills 1165 1165 1140 1130 1120 1110 1110 1000 1000                                                      220°F                                    1090 1010 I

Wet N76 000 lO5 isooj 4 Vapor I 1600 1040 1033IWJ .760 1020 1010 CrlU..1 Pea...., pI. paey. je of Cfl*ir* 140. *1.3. A. Md0 W The yale.. abe.,. *0.1. be.. Ia ayeeeaae ,.iO0. a3 t.l...ar.. .bt.h.d by Lb. Thud 1nthe..I S*_ Tabal. Coal...... .04 a,. d*yad Lye.. the y*ayaeLi.. of Lb. pflO..b e.p.. d..e,bed Th...aodye...h. Pyept .0 pOW by Dr. 10. F. L.ir, of Cocbe.tia E.gloe1n I..., .t Lb. Spd. M..tleg of the A.aiy... So*0ty of M..be.Ie.l Eeooeee., May 1-3, *940, ..ad *.beqe.af1y pbN.h.4 the Tyee...tlo.. t6 *S,M.E., let. 1945.

                                     ,110                 Rp.1ae4 fee.. OT70AM ThrnE$ yCbl1.b.d by COMBUSTION ENOINETO.X3ftO. INC, 700 W*....., Coo.. 00*105

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2009 MNS RO NRC Retake Examina QUESTION 40 2240 KA KA_desc APEOO8 Knowledge of the operational implications of the following concepts as they apply to a Pressurizer Vapor Space Accident: (CFR 41.8 / 41.10 I 45.3) LI Thermodynamics and flow characteristics of open or leaking valves AKI.O1 Unit 1 is in Mode 3 at full temperature and pressure. The crew has entered API1IAI5500IO11 (Pressurizer Pressure Anomalies) due to Pressurizer pressure decreasing very slowly.

Pressurizer pressure is 2150 PSIG
PRT pressure is 2 PSIG Given the above conditions, determine which ONE (1) of the following would indicate a leaking PORV and the state of the fluid in the PORV discharge?

REFERENCE PROVIDED PORV Discharge Temperature State of the Effluent A. 200-240°F Wet Vapor B. 200-240°F Saturated Vapor C. 240-280°F Wet Vapor D. 240-280°F Saturated Vapor Monday, February 08, 2010 Page 91 of 171

FOR REVIEW ONLY DO NOT DISTRIBUTE A

2009 MNS RO NRC Retake Examina QUESTION 40 2 General Discussion Required Reference is steam table. This question is associated with TMI. Per the TMI lesson plan: It was clear from the operators understanding of the PZR PORV discharge temperature and the indications of saturation/superheated fluid in the hot leg, that operator knowledge of thermodynamics needed to be drastically improved. At 0520 the operators obtain a printout of PZR Safety and PORV discharge temperatures showing 232°F and 283°F respectively, but the operators still believe the PORV to be closed. For some time the PORV had been leaking prior to this day. The PORV leakage had been accepted as a normal part of operation (i.e. workaround). The temperature on the discharge of the PZR PORV had indicated what would be seen for PORV open or leaking since the PORV had started leaking. The operators believed the discharge temperature would increase to PZR temperature if the PORV actually opened. A Pressurizer pressure of2150 psig (2165 psia) corresponds to a Saturated Vapor Enthalpy of 1125 BTU/lbm. This Enthalpy undergoing a throttling process discharging to a PRT at a pressure of 2 psig (17 psia) This KA is matched since the applicant must know how to use the Mollier diagram to determine the thermodynamic characteristics of the fluid entering the PRT. This is an analysis level because the applicant must evaluate the given conditions using the Mollier diagram to determine the correct temperature and state of the fluid. Answer A Discussion CORRECT. Answer B Discussion Incorrect. Plausible as this temperature would be obtained if the student followed the entropy line from the PRT pressure to the saturation curve. Answer C Discussion Incorrect. Plausible if the applicant follows the entropy line from the PRT press to the saturation curve. Answer D Discussion Incorrect. Plausible as this temperature would be obtained if they follow the entropy line from the PRT press to the saturation curve. Basis for meeting the KA Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 2006 NRC Q2 (Bank 608) Development References Student References Provided THFFLOO7 Steam Tables OP-CN-II-TMI KA KA_desc APEOO8 Knowledge of the operational implications of the following concepts as they apply to a Pressurizer Vapor Space Accident: (CFR 41.8 /41.10 / 45.3)ElThermodynamics and flow characteristics of open or leaking valves AK1.O1 O1 -9 Comments: RemarkslStatus APEOO8AK1 .01 Considered SAT for submittal with no comments. No changes. Double jeopardy with Q 9: This Q is not double jeopardy with Q 9 because 9 dealt with a ruptured PRT. This Q deals with the usage of the mollier diagram. I am researching the SRO ONLY aspect of this. RFA 10/28/09 Monday, February 08, 2010 Page92ofl7l

FOR REVIEW ONLY DO NOT DISTRIBUTE A

2010 MNS SRO NRC Examination QUESTION 41 254l EPEOO9 EK2.03 - Small Break LOCA ow1edge of the interrelations between the small break LOCA and the following: (CFR 41.7/45.7) iGs Given the following conditions on Unit 1:

The unit has experienced a Reactor Trip and Safety Injection due to a Small-Break LOCA
The crew has just completed the actions of E-0 (Reactor Trip or Safety Injection)
NV pump flow to the NC system Cold Legs is 390 GPM
NC system pressure is 1300 PSIG and stable
SG pressures are 1092 PSIG and stable
NC system subcooling on the ICCM is 22°F and stable Which ONE (1) of the following describes plant conditions upon transition to E-1 (Loss of Reactor or Secondary Coolant)?

SGs NC Pumps Required for Running? Heat Removal? A. YES YES B. YES NO C. NO YES D. NO NO Tuesday, July 13, 2010 Page 116 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 41 L 25411

,aneraI Discussion For this plant condition, even though NV pumps are running and injecting into the NC system, since NC subcooling is not less than 0°F, NC pumps should still be running (E-0 Foldout Page requirement).

Additionally, since NC system pressure is greater than SG pressures and both NC system and SG pressures are stable, the SGs are required for NC system cooling. Answer A Discussion CORRECT: See explanation above. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is correct. Part 2 is plausible because the applicant may conclude that the SGs are not required for NC system heat removal since there is 390 GPM of flow to the cold legs from the NV pumps. Answer C Discussion - -_____________________________________ INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because the applicant might conclude that NC pumps should not be running since a Safety Injection has occurred and the NV pumps are injecting into the cold legs at 390 GPM. However, the NC pumps are only secured in accordance with E-0 Foldout Page criteria if the NV pumps are running and NC system subcooling has been lost. Part 2 is plausible because the applicant may conclude that the SGs are not required for NC system heat removal since there is 390 GPM of flow to the cold legs from the NV pumps. Answer D Discussion INCORRECT: See explanation above.

  .AUSIBLE: Part 1 is plausible because the applicant might conclude that NC pumps should not be running since a Safety Injection has occurred and the NV pumps are injecting into the cold legs at 390 GPM. However, the NC pumps are only secured in accordance with E-0 Foldout Page criteria if the NV pumps are running and NC system subcooling has been lost.

Part 2 is correct. Basis for meeting the KA This KJA is met because the applicant must evaluate a given situation where a small break LOCA has occurred and determine that the SGs are still required for NC system heat removal. Basis for Hi Cog This is a hi cog question because it requires more than one mental step. First the applicant must analyze the given conditions and compare them to recalled memory (E-0 Foldout Criteria) to determine that the NC Pumps should still be running. Additionally, the applicant must analyze the given conditions to determine that with NC system pressure stabilizing above the secondary safety valve set pressure, that break flow is not sufficient to remove all decay heat energy and that the SGs are required for NC system heat removal. Basis for SRO only LZ Lb Level Cognitive Level QuestionType Question Source RO Comprehension BANK VCS Nuclear Station 2007 Audit Examination velopment References Student References Provided WOG HPBG-E-l, Rev 2, Section 2.1, 3/8<break<l, pages 7 & 8

    -MC-EP-E1 Obj 7 EPEOO9 EK2.03 Small Break LOCA
                    -

Knowledge of the interrelations between the small break 1.OCA and the following: (CFR 41.7 / 45.7) S/Gs Tuesday, July 13, 2010 Page 117 of 294

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2010 MNS SRO NRC Examination QUESTION 41 401-9 Comments: RemarkslStatus 401-9 Comments: Distractor D is NP because of the way it is written. One would not say NC pumps are not running. SGs are not required for NC system heat removal. Consider setting up A D in a table

format and ALL distractors will be plausible. Resolution / Comments: Not sure that I completely understood what were looking for here. However, wrote a proposed replacement question with the answers in table format. See attached file for_proposed_revision. Tuesday, July 13, 2010 Page 118 of 294

Question 41

References:

OBJECTIVES s NNLLL E OBJECTIVE L L P P 0 OORSR R0O 1 Explain the purpose for each procedure in the E-1 series. X X EPE1 001 2 Discuss the entry and exit guidance for each procedure in the X X E-1 series. EPE1 002 3 Discuss the mitigating strategy (major actions) of each X X X procedure in the E-1 series. EPE1 003 4 Discuss the basis for any note, caution or step for each X X X procedure in the E-1 series. EPE1 004 5 Given the Foldout page discuss the actions included and the X X X basis for these actions. EPE1 005 6 Given the appropriate procedure, evaluate a given scenario X X X describing accident events and plant conditions to determine any required action and its basis. EPE1 006 7 Discuss the time critical task(s) associated with the E-1 series X X X procedures including the time requirements and the basis for these requirements. EPE1 007

From Lesson Plan OP-MC-EP-E1 Pg. 13 (Rev 23) 1.0 PROCEDURE SERIES BACKGROUND 1.1. E-1, Loss of Reactor or Secondary Coolant 2.1.1 Loss of Reactor Coolant In order to describe the various phenomena that can occur during a LOCA, it is convenient to define five categories of accidents based on the size of the break and number of S/I trains. This section describes four break sizes and Safeguard equipment status as follows: I. Breaks between 3/8 (O.l in ) and 1 (O.8 in 2 ) diameter with minimum safety 2 injection. NC pressure will stabilize above steam generator pressure.

2. Breaks between 3/8 (O.1 in ) and 1 (O.8 in 2 ) diameter with maximum 2

safety injection. The NC will repressurize.

3. Breaks between 1 (O.8 in ) and 13.5 (1 ft 2 ) diameter. NC pressure goes 2

below steam generator pressure.

4. Breaks greater than 1 ft

                                  . The NC will rapidly depressurize to close to the 2

containment atmospheric pressure. Breaks smaller than 3/8 (O.1 in ) with normal charging are considered to be leaks 2 rather than small LOCAs since NC pressure and Pzr level do not go down. If charging flow is not available, the transient would be similar to the response described below for small LOCAs. SMALL LOCAs The flowpath through the E-1 series is dependent upon the break size, the break location, and operator/Station Management decisions. For a break size of up to 1 inch diameter, the amount of S/I flow determines the flow path in the E-1 series. If minimum S/I flow is assumed, the E-1 S/I-termination criteria would not be met, repressurization of the reactor coolant may not occur, and S/I flow equals the break flow. This constitutes a safe and stable condition for the long term provided the heat sink is maintained. As long as S/I and Auxiliary Feedwater are available, the reactor will reach equilibrium conditions for the steam generator pressures. Long-term cooling may require depressurizing to cold shutdown while stepping down S/I flow, so ES-I .2, Post LOCA Cooldown and Depressurization would be used. If maximum S/I flow is assumed such that S/I flow is greater than break flow, the reactor will rapidly repressurize, and may in fact end up with the pressurizer filled solid. At this point, the NC system will rapidly repressurize and the S/I termination criteria will be met, and S/I may be terminated using ES-1.1, S/I Termination. However, if S/I is not terminated, or more realistically, if S/I termination is delayed, the core will remain cooled and in a safe and long term stable condition. The NC system will remain in an acceptable, although possibly not desirable, condition.

From WOG Background Doc HPBG E-1 Sect 2.1 Breaks 3/8 equivalent diameter hole Breaks in this range are considered to be leaks, rather than small LOCAs, since the normal charging system can maintain reactor coolant inventory so that RCS pressure and pressurizer level do not decrease. Very slight system depressurization may occur but no automatic trip or safety injection signal would be generated. The core will remain fully covered provided that the steam generators are available to remove energy, and makeup flow is continuously delivered to the RCS. If charging flow is not available, the RCS transient behavior would be similar to the response described for Category 2. If the leak is within Technical Specification limits or it can be isolated, the plant could remain in power operation. If the leak is above Technical Specification limits and cannot be isolated, then the plant should go to a cold shutdown condition utilizing the normal shutdown procedures. During cooldown the charging system should maintain pressurizer level and the RCS depressurization should be controlled to conform to the normal cooldown limits. Breaks 3/8 <diameter <- 1, minimum safety iniection, or Category 1 breaks above with no charging flow assumed For these break sizes the normal makeup system cannot maintain level and pressure. The RCS will depressurize and an automatic reactor trip and safety injection signal will be generated. Provided that a secondary side heat sink exists, the RCS will reach an equilibrium pressure which corresponds to the pressure at which the liquid phase break flow equals the high pressure pumped safety injection flow. It has been verified that this equilibrium pressure condition will be established for plants with charging/SI pumps. This effect is described here by the presentation of a specific plant analysis for break sizes within this range. A general description of system behavior applicable to the sample transient is provided first, then specific comments concerning the sample analysis are provided.

Early in the transient a loss of subcooled liquid in the RCS occurs which results in a moderate depressurization to the pressure which corresponds to saturation pressure in the core and hot legs. At this point the upper head, upper plenum, hot legs, and core begin to experience some slight voiding, but more than enough liquid flow exists through the core to keep it covered and cooled. During this period of voiding, however, RCS depressurization occurs at a much slower rate than during the time when the entire system was subcooled. Eventually the RCS depressurizes to the point of the reactor trip signal. Immediately following reactor trip, the RCS rapidly depressurizes, since only a fraction of the heat previous to trip is now being transferred to the primary fluid. Due to this rapid depressurization following reactor trip, a safety injection signal is quickly generated. Within a few minutes of the reactor trip time, an equilibrium pressure is established which is above the steam generator pressure. The fluid conditions in the RCS at the time of equilibrium pressure establishment may be characterized by slight voiding in the core and upper plenum and hot legs, and saturated or slightly subcooled liquid in the cold legs. Core heat is removed through the steam generators by continuous single or two-phase natural circulation. The primary mixture level in the steam generators does not drain for breaks of this size, and the core remains covered throughout the entire transient provided that SI is not interrupted. Once equilibrium pressure is established there is no further net loss of liquid volume in the RCS. The natural circulation heat removal mode continues until the time that the break can remove all the decay heat (1 day for a 1 break). Prior to this time, auxiliary feedwater is required to maintain the heat sink. Since the equilibrium pressure established is determined by means of a volume balance of SI flow and break flow, the AP and AT from primary to secondary side, together with the cold safety injection water, may provide a total heat sink greater than the decay heat generated and a cooling of the primary fluid can occur.

Question 41 Parent Question (VCSNS 2007 Audit Examination):

15. 009 EK2.03 1 Given the following plant conditions:

A reactor trip has occurred.
Safety Injection is actuated.
All actions required in EOP-1 .0, Reactor Trip/Safety Injection Actuation, have been taken.
RCS pressure is 1300 psig and stable.
SG pressures are 1050 psig and stable.

Which ONE (1) of the following describes the plant condition upon transition from EOP 1.0? A. RCPs are running. SGs are required for RCS heat removal. B. RCPs are running. SG5 are NOT required for RCS heat removal. C. RCP5 are NOT running. SG5 are required for RCS heat removal. D. RCPs are NOT running. SGs are NOT required for RCS heat removal. A is incorrect. With RCS pressure higher than SG pressure, a secondary heat sink is required. RCPs will be off due to RCS pressure B is incorrect. SG5 are available and required to remove heat from the RCS C is correct. A SBLOCA is in progress as indicated by RCS pressure being above RHR Pump shutoff head. SGs would not be required for heat removal if LBLOCA in progress. RCPs are off D is incorrect. SGs are required Knowledge of the interrelations between the small break LOCA and the following: SIGs. Question Number: RO 41 Tier 1 Group 1 Importance Rating: 3.0 Answer Explanation: ,- Technical

Reference:

Proposed references to be provided to applicants during examination:

Learning Objective: Question Source: Question History: Question Cognitive Level: 10 CFR Part 55 Content: Comments: Answer: C

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 42 APEO 15/017 2.1.32 Reactor Coolant Pump (RCP) Malfunctions

PEO15/017 GENERIC bi1ity to explain and apply system limits and precautions. (CFR: 41.10 / 43.2 /45.12) Unit 1 was operating at 100% RTP. Given the following trends on the 1A NCP: Time 0200 0205 0210 0215 Pump#1 Seal D/P (PSID) 215 210 205 195 Lower pump bearing temp (°F) 221 225 228 231

                 #1 seal outlet temp (°F)                       205        227      235  251 Motorwinding temp (°F)                          312       314       316 323 What is the LATEST time at which the 1A NCP must be secured?

A. 0200 B. 0205 C. 0210 D. 0215 Tuesday, July 13,2010 Page 119 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 42 2542 neral Discussion NCP Trip criteria: Any motor bearing temperature> 195°F Seal Outlet temperature> 23 5°F Motor winding temperature> 311°F (Any bearing water exit temperature> 225°F) 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 the NCP operating limits from the Limits and Precautions. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant does not recall the NCP operating limits from the Limits and Precautions. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible ifthe applicant does not recall the NCP operating limits from the Limits and Precautions. Basis for meeting the KA The K/A IS matched because a malfunction of the 1A NCP has occurred and the applicant must determine based on comparing the given data to the Limits and Precautions for the NCPs when the pump must be stopped. Basis for Hi Cog - This is a higher cognitive level question because it requires more than one mental step. The applicant must first recall from memory the NCP

, 9erating limits from the limits and precautions. The applicant must then analyze the given data to determine when the NCP operating limits are

( eeded. tsis for SRO only .- - - Jo[lgnitive Level QuestionType Question Source RO Comprehension BANK 2003 CNS NRC Q45 Development References Student References Provided sson Plan OP-MC-PS-NCP Lesson Plan Objective 15 APEOI5/017 2.1.32 Reactor Coolant Pump (RCP) Malfunctions

APEOI5/017 GENERIC Ability to explain and apply system limits and precautions. (CFR: 41.10 / 43.2 / 45.12) 401-9 Comments: Remarks!Status osed replacement question for 2010 NRC Q42. Approved replacement_question. RFA 07/07/10 Wednesday, July 14, 2010 Page 120 of 294

Question 42

References:

From Lesson Plan OP-MC-PS-NCP Section 3.1.1: WHEN reactor power greater than 25%, starting an NC Pump is prohibited. BASIS: The concern here is a power excursion which could result in a reactor trip and possible core damage. The idle loop temperature is at Tc for the system and the higher the reactor power the larger the core iT. Objective #15 NC Pump trip criteria are:

Any motor bearing temperature exceeds 195°F.
Any motor winding temperature exceeds 311°F.
The lower pump bearing temperature exceeds 225°F.
The motor frame vibration exceeds 5 mils.
The pump shaft vibration exceeds 20 mils.
The motor shaft vibration exceeds 20 mils.
The flywheel vibration exceeds 20 mils.
The flywheel axial vibration exceeds 20 mils.
High or Low oil level alarm with an adverse trend in either the upper or lower motor oil reservoirs.
No. I seal outlet temperature exceeds 235°F.
ICCM indicates NC System is nearing saturation conditions (loss of subcooling).
The No. I Seal z\P is less than 200 PSI.

BASIS: Stopping a pump when any of these parameters is exceeded should reduce the possibility of any further degradation of the pump or motor. AP/1/A/5500/008 (Reactor Coolant Pump Malfunctions) provides guidance for No. 1 seal leakoff concerns. BASIS: The AP provides the operator with guidance for responding to NCP malfunctions. Starting an NC Pump supplied from the same Auxiliary Transformer through which a DIG is paralleled to the system may result in tripping the D/G breaker. The D/G should be shutdown or the NC Pump transferred to the alternate Auxiliary Transformer before starting.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2003 CNS SRO NRC Examination QUESTION 45 245 APEO15/017 AK2.1O Reactor Coolant Pump (RCP) Malfunctions

Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following: (CFR 41.7 / 45.7) RCP indicators and controls Unit 1 was operating at 100% power. Given the following trends on the 1A NCP: Time 0200 0205 0210 0215 Motor bearing temp (°F) 180 184 186 195 Lower pump bearing temp (°F) 221 225 228 231

              #1 seal outlet temp (°F)                       205     227       235       251 Motor winding temp (°F)                        312     314       316       323 What is the earliest time at which the 1A NCP must be secured?

A. 0200 B. 0205 C. 0210 D. 0215 Wednesday, July 14, 2010 Page 91 of 204

FOR REVIEW ONLY DO NOT DISTRIBUTE A

2003 CNS SRO NRC Examination QUESTION 45 245 General Discussion Bank Question: 870 NCP Trip criteria: Any motor bearing temperature> 195°F Seal Outlet temperature> 235°F Motor winding temperature > 311°F (Any bearing water exit temperature > 225°F) Answer A Discussion Correct: NCP must be stopped if motor winding temperature reaches 311 degrees at 0200 Answer B Discussion Incorrect: NCP must be stopped at 0200 Plausible: reaches the temperature for securing NCP on lower bearing. Answer C Discussion Incorrect: NCP must be stopped at 0200 Plausible: reach the limit for securing NCP on seal outlet temp at 0210 Answer D Discussion Incorrect: NCP must be stopped at 0200 Plausible: reach the temperature for stopping NCP on motor bearing at 0215 Basis for meeting the KA Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK Development References Student References Provided Lesson Plan Objective: NCP Obj: 12

References:

1. OP-CN-PS-NCP pages 7, 10, 14, 15 APEO15/017 AK2.10 Reactor Coolant Pump (RCP) Malfunctions

Knowledge of the interrelations between the Reactor Coolant Pump Malfunctions (Loss of RC Flow) and the following: (CFR 41.7/45.7) RCP indicators and controls 401-9 Comments: Remarks/Status Wednesday, July 14, 2010 Page 92 of 204

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 43 2543 APEO22 AA1.09 Loss of Reactor Coolant Makeup

bility to operate and/or monitor the following as they apply to the Loss of Reactor Coolant Makeup: (CFR 41.7/45.5/45.6) ,.CP seal flows, temperatures, pressures, and vibrations A loss of all charging and seal injection flow on Unit 1 has resulted in a failure of the lB NCP #2 Seal. The lB NCP #1 Seal Leak-off flow is going (1) lB NCP #2 Seal Standpipe (2) level alarm is LIT. Which ONE (1) of the following completes the statements above? A. 1. DOWN

2. LOW B. 1. DOWN

2. HIGH C. 1.UP

2. HIGH D. 1.UP

2. LOW Tuesday, July 13, 2010 Page 121 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 43 2543 B neraI Discussion From the Background Document for AP-08 (Malfunction of NC Pump): If the #2 seal failure is the initial failure on the NC Pump, it would cause a high standpipe level and low flow on #1 seal leakoff. - __________ Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant does not understand the flow path through the NC pump seals and the effect of various seal malfunctions on indicated seal flows and standpipe level. Answer B Discussion CORRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant does not understand the flow path through the NC pump seals and the effect of various seal malfunctions on indicated seal flows and standpipe level. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant does not understand the flow path through the NC pump seals and the effect of various seal malfunctions on indicated seal flows and standpipe level. Basis for meeting the KA -_________________ The KA is matched because Reactor Coolant Makeup is lost and subsequently restored resulting in a malfunction of the lB NCP #2 seal. The applicant demonstrates the ability to monitor RCP seal flows by demonstrating a knowledge of what indications would confirm a failure of the

 #2  Seal.

Basis for Hi Cog is a higher cognitive level question because it requires multiple mental steps. The applicant recall from memory the flowpath though the #1 flisd #2 Seals and determine the impact of a number #2 Seal failing opening on the seal flow indications. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK MNS Exam Bank Question #PSNCPNO4 Development References Student References Provided Learning Objectives:

1) PS-NCP #12

References:

1) Lesson Plan OP-MC-PS-NCP Section 2.3.2

2) AP-08 Background Document APEO22 AA1 .09 Loss of Reactor Coolant Makeup

Ability to operate and! or monitor the following as they apply to the Loss of Reactor Coolant Makeup: (CFR 41.7/45.5 /45.6) RCP seal flows, temperatures, pressures, and vibrations 401-9 Comments: Remarks!Status 40 1-9 Comments: The answer choices are very difficult to read. Consider setting this Q up as a fill in the blank as follows: The #1 Seal Leak off Flow is going NC Pump number 2 Seal Standpipe level alarm is LIT A. Down Low Tuesday, July 13, 2010 Page 122 of 294

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2010 MNS SRO NRC Examination QUESTION 43 B. Down High C. Up High D. Up Low Resolution I Comments: Revised question per Lead Examiners recommendation. See

attathed file for revised question.

Tuesday, July 13, 2010 Page 123 of 294

Question 43

References:

From Lesson Plan OP-MC-PS-NCP Section 23.2: Objective #11,12 Each of the NC Pump No. 1 seal leakoff lines have seal return isolation valves. These valves are closed when NC System pressure is less than 100 psig in order to prevent any backflow from the NV System through the seal return filter to the NC Pump seals. Backflow would flush any contaminants/particulates out of the filter and into the seal. These isolation valves are also used in the event of a failure (excessive leakage) of the No. 1 seal. When No. I seal leakoff flow is high, some of this flow comes from the NC System up through the thermal barrier. There may be insufficient heat removal by the thermal barrier heat exchanger to adequately cool the leakoff flow. This hotter water could cause damage to the No. 2 and 3 seals. When the seal return valve is closed, the No. 2 seal becomes the primary seal and maintains the large AP. The No. 2 seal is designed to withstand this high AP for a short period and the pump must be stopped within 5 minutes (per ESBU-TB-93-01-R1) and the plant must be cooled down and depressurized so that repairs can be made. The NC Pumps are equipped with a common No. 1 seal bypass valve. This valve is only opened at low system pressures (100-1000 psig) when there is insufficient flow to adequately cool the seal (leakoff temperature >200°F). The leakoff from each pump is piped to a common manifold and then via a seal water filter through a seal water heat exchanger where the temperature is reduced to about that of the VCT. Leakage past the No. 1 seal provides a constant pressure on the No. 2 seal and constant pressure on the No. 3 seal. A standpipe is provided to assure a backpressure of at least 7 feet of water on the No. 3 seal. In addition, the standpipe is used to warn of excessive No. 2 seal leakage flow to the reactor coolant drain tank (NCDT). Excessive No. 2 seal leakage results in a rise in the standpipe level and eventual overflow to the NCDT via a second overflow connection. A total of 8 gpm is supplied to each NC pump for seal injection water. 5 gpm is directed down through the thermal barrier labyrinth seal and into the NC System. 3 gpm flows up through the lower radial bearing. A minimum differential pressure of 200 psid is required at low NC System pressure (see 7.6) across the No. 1 seal surfaces to ensure proper water film during pump operation. For an NCP start at normal system pressure, there must be approximately 1 gpm seal leakoff flow for 2200 psid. The inlet pressure is approximately 2250 psig (NC System pressure) and the outlet pressure is 15-50 psig (VCT pressure) during normal operation. Approximately 3 gpm leaks off from the No. 1 seal of which 3 gph flows to the No. 2 seal. Proper VCT pressure is required to ensure adequate backpressure for proper flow through the No. 2 seal. Objective #12 I Approximately 3 gph is directed through the No. 2 seal. The pressure drops from 50 psig to 3 psig across this seal. All the No. 2 seal leakoff, except for 100 cc/hr, is directed to a standpipe. The water level in the standpipe is maintained to provide

sufficient backpressure on the No. 2 seal to ensure flow through the No. 3 seal. All excess water from the standpipe is discharged to the NCDT through an orifice. Improper standpipe level can adversely affect seal operation, therefore there is a high and low level alarm provided for the standpipe to warn of potential seal problems. A high level alarm could indicate excessive No. 2 seal leak-off flow. Approximately 100 cc/hr from the No. 2 seal is directed to the No. 3 seal. The pressure drops from 3 psig to atmospheric across this seal. After passing through the seal the leakoff is directed to the NCDT. The minimum and maximum flow rates and temperatures for seal injection water are 6 gpm and 50° F and 12 gpm and 150° F, respectively. I Objective #9 No.1 seal temperature, injection flow, and AP indications are provided on the Main Control Board. Recorders are provided for No. 1 seal leakoff flow indicating low range (0-2 gpm) and high range (0-6 gpm) flow. Other indications are provided on the OAC. 2.4 NC Pump Monitor System Objective #15 The purpose of the EME system is to monitor the voltage and frequency of the 6900V power source for the reactor coolant pump motors. Following a drop in either parameter below its setpoint, the monitoring system will provide a signal to the Solid State Protection System (SSPS) to indicate the condition. Due to the direct impact of the EME system on the performance of the Reactor Protection System (through the SSPS reactor trip circuit), it is classified as nuclear safety related. By definition, the Reactor Protection System is designed to shut down the reactor to protect against fuel cladding damage or loss of system integrity, which may result in the release of radioactive fission products into Containment. The under-voltage and under-frequency monitors are voltage and frequency sensing devices, respectively. Frequency is monitored between the supply breaker and the safety breaker while voltage is monitored between the safety breaker and the motor (see drawing 7.16). Each monitors output sends a signal to its corresponding auxiliary relay which in turn sends a signal to the SSPS to indicate the condition. If 2 out of the 4 channels monitored indicate an under-voltage (or under-frequency) condition, the SSPS will initiate a reactor trip (1/4 causes an NC Pump Bus Alert alarm in the Control Room). The underfrequency and undervoltage setpoints are listed in Technical Specifications table 3.3.1-1. For an under-frequency condition, the reactor coolant pump circuit breakers will be tripped as well. Note that at a power level less than the P-7 interlock the SSPS Reactor Trip Function on under-voltage and/or under-frequency will be blocked. However, the P-7 interlock will NOT block the trip of the Reactor Coolant Pump Safety Breakers. If an under-frequency condition exists on 2 or more pumps, the Reactor Coolant Pump Safety breakers will be tripped, regardless of the power level.

From AP-08 Background Document: DISCUSSION: An observed NC Pump seal phenomenon following seal repair or replacement is that a short period of operation (up to 24 hours) may be required to get the seals to seat properly. This can result in abnormal seal leak-off. For example, if the #1 seal is not seated properly, it can cause the #1 seal leak-off to be high. If the #2 seal is not seated properly, it can cause the #1 seal leak-off to be low (with #2 seal standpipe high level alarm). This note is for consideration only. The following steps are not affected by this information. If leak-off is not within the normal range (1 .0 gpm 5.0 gpm), but within the

pump operating limits, direction is given to contact station management for further guidance and continue to monitor NC Pump seal leak-off flow. Continued NC Pump operation is allowed within the operating limits (0.8 gpm 6.0 gpm). The information in

this note would be taken under consideration by station management.

REFERENCES:

MCM 1201.01-0193 001, NCP Instruction Manual, Section 6.5Troubleshooting

CASE I STEP 17: PURPOSE: Diagnose a #2 seal failure during scenarios where #1 seal has not failed. DISCUSSION: This point in the AP is reached with NC Pumps running. Possibly the only failure is the

2 seal. Typically, as the #2 seal fails, its leak-off increases. Two effects of its leak-off increasing are #1 seal leak-off decreasing and standpipe level increasing. This step checks for these symptoms and if true, #2 seal failure is assumed. Direction is given to continue to monitor pump parameters, notify Engineering to determine #2 seal leakoff flow, and evaluate continued NCP operation. Revision 1 to Westinghouse Product Update S-013 provides an operational limit of 1.1 gpm for #2 seal leakoff flow.

Reference PIP M-08-5384. The Engineering notification was discussed with Steve Rosenau, NCP component Engineer.

REFERENCES:

PIP M-08-5384 Westinghouse Product Update S-013 Rev. 1

Question 43 Parent Question: Question 72 PSNCPN04 PSNCPNO4 1 Pt Unit 1 is at 100% power when indications are received of a iB Reactor Coolant Pump seal malfunction. AP/1/A15500/008 (Malfunction of NC Pump) is implemented. Which one of the following conditions describes a number two seal failure? A. * #1 Seal Leak off flow GOING DOWN

NC Pump number 2 Seal Standpipe low level alarm - LIT
NCDT input STABLE, OR GOING DOWN

B. * #1 Seal Leak off flow GOING DOWN

NC Pump number 2 Seal Standpipe high level alarm - LIT
NCDT input GOING UP

C. * #1 Seal Leak off flow GOING UP

NC Pump number 2 Seal Standpipe high level alarm - LIT
NCDT input STABLE, OR GOING DOWN

D. * #1 Seal Leak off flow GOING UP

NC Pump number 2 Seal Standpipe low level alarm - LIT
NCDT input GOING UP

Answer 72 B Distracter analysis: A. Incorrect: Plausible: # 1 Seal LIO WILL go down C. Correct C. Incorrect: Plausible: High Standpipe level alarm WILL light D. Incorrect: Plausible: NCDT input WILL go up Level: RO & SRO KA: SYS 003 (3.1 /3.0) Lesson plan objective: OP-MC-PS-NCP, Obj 12 Source: New Level of knowledge: Comprehension

Reference:

1. OP-MC-PS-NCP, pgs 25-29

2. AP/1/N5500/008, Malfunction of NC pump

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 44 APEO25 AA1.12 Loss of Residual Heat Removal System (RHRS)

( .bility to operate and! or monitor the following as they apply to the Loss of Residual Heat Removal System: (CFR 41.7 / 45.5 /45.6) CS temperature indicators Given the following conditions on Unit 1:

Unit is in Mode 5
Both Trains of ND are initially in service
NC system temperature is being maintained at 140°F
Subsequently, both ND pumps trip
The crew has implemented AP-19 (LOSS OF ND OR ND SYSTEM LEAKAGE)
Efforts to restore an ND pump to service have been unsuccessful If a MAXIMUM NC system temperature of (1) is exceeded, AP-19 will direct the crew to stop attempts to restore an ND pump and (2) to restore cooling to the NC system.

Which ONE (1) of the following completes the statement above? A. 1. 180°F

2. initiate NC system feed and bleed B. 1. 180°F

2. attempttostartan NC pump C. 1. 212°F

2. initiate NC system feed and bleed D. 1. 212°F

2. attempt to start an NC pump Tuesday, July 13, 2010 Page 124 of 294

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2010 MNS SRO NRC Examination QUESTION 44 2544 neral Discussion JAW AP-19 Step 19.d. if NC system temperature exceeds 180°F then NC system feed and bleed is initiated. Answer A Discussion CORRECT: See explanation above. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is correct. Part 2 is plausible because this is an option in AP-19 if there is still a bubble in the Pressurizer. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because throughout AP-19 the term saturated conditions is reference frequently. 212 deg is normally associated with water being saturated. Part 2 is correct. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because throughout AP-19 the term saturated conditions is reference frequently. 212 deg is normally associated with water being saturated. Part 2 is plausible because this is an option in AP-19 if there is still a bubble in the Pressurizer. Basis for meeting the KA e KA is matched because the applicant demonstrates an ability to monitor NC system temperature during a loss of RHR by demonstrating a 3wledge of the temperature at which compensatory actions must be taken in AP-19. basis for Hi Cog Basis for SRO only Job LeveI Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Learning Objective:

1) AP-19 #3

References:

1) AP-19 Loss of ND or ND System Leakage APEO25 AA1.12 Loss of Residual Heat Removal System (RHRS)

Ability to operate and! or monitor the following as they apply to the Loss of Residual Heat Removal System: (CFR 41.7! 45,5 / 45.6) RCS temperature indicators 401-9 Comments: RemarksIStatus L_________________________ 401-9 Comments: No comment. Resolution / Comments: Tuesday, July 13, 2010 Page 125 of 294

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2010 MNS SRO NRC Examination QUESTION__44 Tuesday, July 13, 2010 Page 126 of 294

Question 44

References:

From AP-19: MNS LOSS OF ND OR ND SYSTEM LEAKAGE PAGE NO. AP/21A5500/19 18 ot217 Rev. 20 UNIT 2 19. follows: to AOTION/EX?ECTEP RE5FONSE Determine time reach saturation as

a. Check core exit T/Cs AVAILABLE,

F.ESONSE NOT OTAiNEt

a. Perform the following:

1) Estimate time to reach saturation using THERMAL MARGIN (on 2MC-6).

2) GOTOStepl9.c.

b. Monitor temperature heat up and estimate time to reach saturation.

c. Check any NV pump or NI pump - c. Perform the following:

FUNCTIONAL.

1) IF 2FW-27A (FWST Supply To ND) is energized, THEN GO TO Step 19.d.

V 2) IF AT ANY TIME either of the following conditions is met, THEN initiate feed and bleed PER Steps 20 and 21:

                                                               . NC temperature greater than 180F.

OR

                                                               . Core within 20 minutes of reaching saturation.

_3) OOStep22.

d. Check NC temperature - d. Perform the following:

GREATER THAN 180F. 1) 1FATANY]iMEthetemperatureor time conditions are met, THEN OR initiate feed and bleed PER Steps 20 and 21.

          . CORE LESS THAN 1OMINUTESOF REACHING SATURATION.                         2) GO ID Step 22.

Jr

MNS LOSS OF ND OR ND SYSTEM LEAKAGE PAGE NO. APJ2/A/5500/1 9 19 of 217 Rev. 20 UNIT 2 A:TI0N/ExPECTEO ESPCNSE FE5IONSE NOT OBTAINED

20. PrIor to establishing feed and bleed In next step, perform the following:

Ensure any personnel inside NC piping (including SfGs) are outside piping.
IF time allows, prior to NC System boiling, THEN ensure all personnel are evacuated from lower containment.
IF time allows, THEN refer to Data Book curve 2.10.4 (Core Flow Required to Prevent Boiling for Loss of Decay Heat Removal) to estimate minimum feed flow required to stabilize NC temperature.
Announce the following on plant page:

a. Initiating Unit 2 NC System Feed and Bleed.

b. All personnel evacuate Unit 2 containment.

CAUTION

While boIling exists, RVLIS is the preferred NC level indication if available. Flow through the Pzr surge lIne may cause other NC level Indications to be erroneously high. BoilIng may make ultrasonIc level indicatIon unreliable. The following formula converts %UR RVLIS to WR level:
WR level (inches above centerline) = (%UR RVLIS 66) x 4.94

If NC System Slghtglass in service, then WR and NR NC System Level indication will become invalid at 146 duo to spillover into reference legs.

21. Establish NC feed and bleed as follows:

a. Check power to all Pzr PORV isolation a. Evaluate cause of power loss and valves AVAILABLE,

                  -                                            initiate actions to restore power to affected isolation valve(s).

b. OPEN all PZR PORV isolation valves.

c. OPEN all PZR PORVs.

FOR REVIEW ONLY DO NOT DISTRIBUTE B

2010 MNS SRO NRC Examination QUESTION 45 2545 APEO27 AK2.03 Pressurizer Pressure Control System (PZR PCS) Malfunction

nowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: (CFR 41.7/45.7) ..ontrollers and positioners Based on the current Pressurizer Pressure Channel indications below: Pressurizer Pressure Channels Which ONE (1) of the following lists the Selected pressure output to the Pressurizer Pressure Master Controller:

1. Based on the pressure indications shown above?

2. If Pressurizer Pressure Channel 3 fails low?

A. 1. 2240 PSIG

2. 2230 PSIG B. 1. 2235 PSIG

2. 2232.5 PSIG C. 1. 2240 PSIG

2. 2232.5 PSIG D. 1. 2235 PSIG

2. 2230 PSIG Tuesday, July 13, 2010 Page 127 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 45

   .neraI Discussion rThe Pressure Control Signals are developed using a Median Select Second Highest Algorithm receiving input from the available pressurizer pressure channels. Each of the 4 median select logics will provide the median (middle) of the three pressure channel inputs as its output to the high channel select. The high select will then select the highest of the median select channel outputs as the selected channel. Provided there are no problems with the pressure input channels, the selected pressure channel for control will always be the second highest reading channel.

With the conditions given in this question Channel 1 is initially the second highest reading channel. So the selected pressure is 2235 PSIG. When Pressure Channel 3 fails low, the 3 Median select channels with input from Channel 3 will now average the remaining two inputs to provide an output to the High Median Select. Therefore the Medial Select outputs will be as follows: 1 - 2232.5 PSIG 2 - 2230 PSIG 3 - 2227.5 PSIG 4 - 2230 PSIG The highest Median Select Channel is now Channel 1 so the output the Pressurizer Pressure Master Controller will be 2232.5 PSIG. Answer A Discussion - INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible if the applicant concludes that the algorithm selects the highest reading channel instead of the highest reading MEDIAN channel. Part 2 is plausible if the applicant concludes that the algorithm choses the MEDIAN channel after the failure instead of the HIGHEST MEDIAN channel. -_________________ Answer B Discussion jCORRECT: See explanation above. iswer C Discussion JCORRECT: See explanation above. 1 PLAUSIBLE: Part 1 is plausible if the applicant concludes that the algorithm selects the highest reading channel instead of the highest reading MEDIAN channel. Part 2 is correct. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is correct. Part 2 is plausible if the applicant concludes that the algorithm choses the MEDIAN channel after the failure instead of the HIGHEST MEDIAN channel. Basis for meeting the KA The KA is matched because the applicant must determine the effect of a pressurizer pressure channel failure on the output of the median select circuit to the Pressurizer Pressure Master Controller. Basis for Hi Cog This is a higher cognitive level question because it requires more than one mental step to arrive at the correct answer. The applicant must first recall from memory that the channel select circuit is a Median Select 2nd Highest Algorithm. Then the applicant must compare the channel indications to determine which of the indications is the second highest indication. After the channel failure (in this case the highest reading channel) the applicant must compare the remaining channels and calculate the average of the remaining two pressure channels (for the median select channels which have input from the failed channel) to determine the second highest channel. Basis for SRO only Job Level Cognitive Level QuesonType Question Source RO Comprehension NEW Tuesday, July 13, 2010 Page 128 of 294

________ FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 45 2545 Oevelopment References Student References Provided rning Objectives: j PS-IPE #3

References:

1) Lesson Plan OP-MC-PS-IPE Section 2.2 APEO27 AK2.03 Pressurizer Pressure Control System (PZR PCS) Malfunction

Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: (CFR 41.7 / 45.7) Controllers and positioners 401-9 Comments: RemarksIStatus 40 1-9 Comments: Why would the applicant select 2230 psig based on this logic? I dont believe Al and Cl are plausible. Facility please explain. This Q is U until facility explains why Al and Cl are plausible. Resolution I Comments: There are better choices for plausibility. Decided to change 2230 PSIG to 2240 PSIG and 2227.5 PSIG to 2230 P51G. Since the algorithm picks the highest median select signal it would be plausible for the applicant to pick the highest reading channel as the output to the Master Controller. See attached file for posed_revision. Need to work on distracter analysis. Tuesday, July 13, 2010 Page 129 of 294

Question 45

References:

From Lesson Plan OP-MC-PS-IPE Section 2.2: 2.0 COMPONENT DESCRIPTION 2.1 Pressurizer Pressure Channels Objective #3 The pressurizer pressure transmitters are also called narrow range pressure transmitters as they span a range of pressure from 1700 to 2500 psig. This range encompasses all the setpoints for control and protective actions that need be taken for power operating conditions. The pressure transmitters (channels 1 through 4) tap off the wet reference legs of the pressurizer level transmitters channels I through 3. ((See drawing 7.3 Pressurizer Pressure and Level Indication (11/26/08) for specifics.)) Each channel is displayed on the MCB, with CH I also displayed on the Auxiliary Shutdown Panel (ASP). Most control and alarm functions are normally provided from Selected Pressurizer Pressure 1 or Selected Pressurizer Pressure 2. The pressurizer pressure control signals are developed using a Median Select Second Highest algorithm. The Selected pressurizer pressure signal is displayed on the pressurizer pressure recorder. 2.2 Pressurizer Pressure Control Signals Objective #3 Refer to Drawing 7.3, Composite Pressurizer Pressure Control. The Pressure Control Signals are developed using a Median Select Second Highest Algorithms receiving input from the available pressurizer pressure channels. Selected Pressurizer Pressure 1, inputs to the Pressurizer Master Controller (heaters, sprays, Low/Hi Press Dev. Annunciators, & PORV NC 34A), the MCB Recorder, and the Low Pressure Interlock for PORVs NC-32B and NC 36B(2185 psig). Selected Pressurizer Pressure 2, inputs the pressure signal to PORVs NC 32B and NC-36B (lift setpoint) 2335 psig, the High pressure alarm (setpoint 2310 psig) and the Low Pressure Interlock for NC-34A (setpoint 2185 psig). 2.3 Pressurizer Pressure Master Controller The Pressurizer Pressure Master Controller (Soft Panel Only) compares actual pressure (Median Select 2nd Highest) with a reference pressure. The reference pressure is entered on the graphic soft controller. Refer to Drawing 7.13, PZR Pressure Control DCS Graphic. Using the PZR PRESS MASTER Pop-up on the PZR Pressure Control Graphic, the operator will depress the A button and using the Increase/Decrease pushbuttons underneath can adjust the setpoint to the desired value. The range of the Master controller is 1700 to 2335 psig with the normal setpoint being 2235 psig. The difference between actual pressure and reference pressure generates a pressure error. Depending on the size and polarity of the error, the Pressurizer Pressure Master will cause various control functions to actuate in attempts to restore actual pressure back to the reference value.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 46 APEO4O AA2.03 Steam Line Rupture

bility to determine and interpret the following as they apply to the Steam Line Rupture: (CFR: 43.5 / 45.13) \. jifference between steam line rupture and LOCA Given the following conditions on Unit 1:

Pzr level is slowly decreasing
Charging flow is slowly increasing
NC system temperature is decreasing slowly
S/G levels are being controlled at program level
The Rod Control Power Mismatch (PMM) indication is (+)1 .5°F Which ONE (1) of the following describes the procedure that will be entered and the FIRST action required based on current conditions?

A. API1IAI5500IOO1 (Steam Leak) Reduce turbine load to maintain Rx power less than or equal to 100%. B. APIIIAI5500IOOI (Steam Leak) Manually throttlel NV-238 (Charging Line Flow Control) to stabilize Pzr level. n C. AP/1/A/5500101 0 (Reactor Coolant Leak) Case II (NC System Leak) Reduce turbine load to maintain Rx power less than or equal to 100%. D. AP/1/A/5500/01 0 (Reactor Coolant Leak) Case II (NC System Leak) Manually throttle 1 NV-238 (Charging Line Flow Control) to stabilize Pzr level. Tuesday, July 13, 2010 Page 130 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 46 LJ A neraI Discussion The stem of the question provided indications, most of which are common to both a small LOCA and a small steam break. The applicant is asked to select the correct procedure by identiiiing the event. The indication given that is not consistent with a LOCA is elevated reactor power. The actions given to either reduce reactor power or manually throttle charging flow are addressed in both procedures but it the scenario given, the first action to be performed by the operators would be to reduce turbine load. - Answer A Discussion CORRECT: See explanation above. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Procedue is correct and the actions provided are contained in AP-Ol but come later in the procedure. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: PZR level and pressure are correct for an NCS leak, but Tave and power would not be affected. Response is correct to address the overpower condition. Answer D Discussion - INCORRECT: See explanation above. PLAUSIBLE: PZR level and pressure are correct for an NCS leak, but Tave and power would not be affected. Response is correct if the event was a LOCA. Basis for meeting the KA K/A is matched because in order to correctly answer this question, the applicant must demonstrate the ability to interpret a given set of conditions and determine whether the cause is due to a steam break or LOCA. Basis for Hi Cog ______________________ This is a hi cog question because it involves a level of analysis of given situation. This involves a multi-part mental process where the applicant ist evaluate the indications given and determine its meaning related to the scenario given. dasis for SRO only - Job Level

         --             Cognitive Level           QuestionType                                       Question Source RO                Comprehension                  BANK                                          Bank CNS Q878 Development References                                                                           Student References Provided From AP-0 1 Background Document Pg 2 From AP-Ol Background Document Pg 4 E]From AP-Ol (Steam Leak) Pg 3 of 42 Li From AP-10 Case 2 (Steam Leak) Pg 38 of 127 OP-MC-TA-AT Obj. 05 OP-MC-AP-1 Obj. 2 APEO4O AA2.03 Steam Line Rupture
                     -

Ability to determine and interpret the following as they apply to the Steam Line Rupture: (CFR: 43.5 /45.13) Difference between steam line rupture and LOCA 401-9 Comments: RemarkslStatus 401-9 Comments: Its common knowledge that power would not go up for a LOCA. Drop the first bullet and C and D will be plausible. The Q is currently a U because C and D are NP. Resolution / Comments: Tuesday, July 13, 2010 Page 131 of 294

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2010 MNS SRO NRC Examination QUESTION 46 By removing power completely from the question, the applicant has no way to discriminate whether the event is a Steam Leak or a LOCA. As a compromise, removed power indication and pressurizer pressure indication bullets and added a bullet providing Rod Control system Power Mismatch (PMM). By doing this the applicant is not given power indication directly and has to interpret the Rod Control system indication to determine that Nuclear Power is increasing faster than Turbine Power which would indicate a steam leak. Also, replaced should with will in the stem of the question per Lead Examiners General Comments. See attached file for proposed question revision. Tuesday, July 13,2010 Page 132 of 294

Question 46

References:

OP-MC-TA-AT Obj. 05 s NNLLL OBJECTIVE L L P P 0 E OORSR R00 4 Given the initial conditions, discuss Abnormal Transients X X associated with High or Low Failure of an Instrument Channel. TAATOO4 5 Given the initial conditions, discuss Abnormal Transients X X associated with accidents which could occur at McGuire. TAATOO5 6 With the aid of Abnormal and Emergency Procedures, discuss X X the affect the above transients will have on plant normal and emergency systems. TAATOO6

OP-MC-AP-1 Obj. 2 OBJECTIVES NNLLL OBJECTIVE L L P P 0 OORSR R00 Explain the purpose for AP/Ol (Steam Leak). X X X AP01 001 2 Analyze the mitigating strategy (major actions) contained in x x x the procedure. APO1 002 3 Given scenarios describing accident events and plant x x x conditions, evaluate the basis for any caution, note, or step. APO1 003 Given scenarios describing accident events and plant x x x conditions, evaluate conditions which require application of continuous action steps. APO1 004

From AP-Ol Background Document Pg 2 Summary For relatively small steam breaks, normal plant control systems are capable of maintaining nominal or near nominal operating conditions. For a small steamline break upstream of the turbine stop valves, the system transient response would be similar to a step load increase. The secondary system would indicate an increase in load with a resultant decrease in primary system average temperature and pressure. The control rods would withdraw from the core in an effort to restore the primary average temperature if the rod control system was in an automatic mode of operation. Due to the apparent increased load, steam flow from the steam generators would be increasing. With the MSIVs open, all loops would experience increased steam flow. Due to the increased steam flow, the feedwater control valves would modulate to a more open position in an attempt to maintain steam generator water level. As a result, main feed flow would be increased. Another indication of this type of break would be a decreasing water level in the condenser hotwell. A containment temperature and/or pressure increase may be observed if the break occurred inside containment. If the break was outside containment, an audible or visual confirmation of the break may be possible. A drop in generator MW output may also be observed. Larger size breaks may require reactor trip and/or safety injection. A different set of symptoms might be encountered for steam leaks that occur downstream of the turbine (on extraction lines, MSRs, and feedwater heaters). For these locations, it may be possible to observe a change in plant efficiency; however, an audible or visual indication may be the first symptom encountered. ENTRY CONDITIONS This procedure can be entered any time the listed symptoms are encountered. It should be noted that the symptom Observed secondary steam leak is the only symptom that definitively identifies a steam leak (and even then the magnitude of the leak may be considered for entry conditions). The other symptoms could indicate a steam leak, or some other event. In some cases the combination of symptoms can be the best indication the event is a steam leak and not some other event.

From AP-Ol Background Document Pg 4 STEP 2: PURPOSE: Prevent exceeding maximum thermal output and prevent an uncontrolled cooldown. DISCUSSION: Reactivity management dictates controlling reactor power less than or equal to 100%. Since steam demand determines reactor power, the increase in steam demand from the leak must be promptly compensated for by a decrease in steam demand from turbine load. During a transient, if reactor power is less than secondary power, temperature will decrease, adding positive reactivity. This will continue as long as reactor power is lower. The first part of the step (reactor power less than 100%) ensures maximum thermal power is not exceeded. Determining reactor power less than 100% can be difficult during a steam leak transient. Thermal power best estimate is averaged over time, so there is a delay indicating reactor power has gone above 100%. The steam leak cools T-ave, which decreases the flux the excore Nis see, causing them to read potentially several percent low. One good real time indicator of reactor power is the NC loop DJTs. The third part of the step (T-ave at T-ref) ensures the power transient is turned. If turbine load is cut to the extent that T-ave is restored to T-Ref, this indicates that reactor power has caught up with secondary power (or else T-ave couldnt be increasing). Reactor power catches up when the turbine load has been reduced by the amount of the steam leak. The T-ave AT T-REF criterion is also included in this step for those scenarios involving a steam leak with the plant less than 100% power. Again, ensuring T-ave turned is a good indicator enough turbine load has been cut to control the reactivity transient. This step is early in the procedure to prevent unnecessary isolation of LID if NCS inventory can be maintained after reducing turbine load (by not allowing T-ave to continue to decrease).

From AP-Ol (Steam Leak) Pg 3 of 42 A;TIo:T::zrsDTED RESPONSE RESCfl3E NOT O3AflTED C. Operator Actions

1. Monitor Foldout page.

2. Reduce turbine load to maintain the following:

        . E core Nis LESS THAN OR EQJAL
                        -

TO 100%

        . NC Loop Dirs LESS THAN CF D:T
                           -
        . T-Avg  - Al T-REFI

3. Check containment entry IN

                                  -
                                               .QTQ StepS.

PROGRESS.

4. Check steam leak KNOWN TO BE

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

a. Announce All personnel evacuate Unit 1 containment.

b , Actuate the contain ment evacuation alarm.

c. REFER TO RPOik5700101 1 (Conducting a Site Assembly. Site EvacUation, or Containment Evacuation: as time allows.

5. Check Pzr pressure prior to event - IF AT ANY TIME an S/I occurs due to GREATER THAN P11(1955 PSIG). steam leak. THEN GO TO Enclosure 2 (S/I Actions For Steam Break In Modes 3 and 4).

From AP-lO Case 2 (Steam Leak) Pg 38 of 127 MNb NC SYSTEM LEAKAGE WITHIN THE CAPACITY OF BOTH N PUMPo .-, P.A,ENO. API1/435E00/1C 35 of 127 Caseu UNIT 1 NC Sjsteni Leakage Rev. 21 ACTICN/ZZPECIED REESONSE RESPCNSE rci DETADrED

2. Check Pzr level STABLE OR GOING UP.

                        -                            Perform the following as required to maintain level:

a. Maintain charging flow less than 200 GPM at all times in subsequent steps.

ii Ensure 1NV-236 (Charging Line Flow Control) opening.

c. Open 1NV-241 (UI Seal Water lnj Flow Control) while maintaining NC pump seal flow greater than 6 5PM.

d. Reduce or isolate letdown.

e. Start additional NV pump.

f. IF P:r level cannot be maintained greater than 4%. OR P:r level going down with maximum charging flow.

THEN perform the following:

1) IF in mode 3 or above, prior to CLA isolation, THEN perform the following:

a Trip reactor. hi WHEN reactor tripped OR auto S/I setpoint reached. THEN ensure 5)1 initiated. c) GO TO ER1.iA5000/E-0 iReactor Trip or Safety Injectioni. 21 IF in mode 3 after CLA isolation or in mode 4. THEN GO TO API k550Oi34 Shutdown LOCAl.

3) IF T-Avg is less than 200°F, THEN GO TO API /AI5500)1 9 :Loss Of ND Or ND System Leakage:.

while continuing in this procedure as time allows.

Parent Question CNS Q878 CNS RO Q878 Ans: C Unit 1 is at 100% RTP. The RO reports the following plant parameters:

RTP has slowly increased to 100.2%
Pzr Pressure is slowly decreasing
Pzr Level is slowly decreasing
Charging Flow is slowly increasing
NC Temperature is slowly decreasing
SIC Levels are being controlled at program level Which one of the following describes the procedure that should be entered and the FIRST action required based on current conditions?

A. AP/1/A/5500/010 (Reactor Coolant Leak) Case II (NC System Leak) Reduce turbine load to maintain Rx power less than or equal to 100% B. AP/1/A15500/010 (Reactor Coolant Leak) Case II (NC System Leak) Manually throttle I NV-294 (NV Pmps A&B Disch Flow Ctrl) to stabilize Pzr level. C. AP111A155001028 (Secondary Steam Leak) Reduce turbine load to maintain Rx power less than or equal to 100% D. AP111A155001028 (Secondary Steam Leak) Manually throttle 1 NV-294 (NV Pmps A&B Disch Flow Ctrl) to stabilize Pzr level.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 47 2547 APEO54 AK1.02 Loss of Main Feedwater (MFW)

nowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater(MFW): (CFR41.8/41.1O I 5.3) Effects of feedwater introduction on dry SIG Given the following conditions on Unit 1:

The unit has experienced a feedwater line break of the 1A SIG inside containment and a total loss of feedwater
FR-H.1 (Response to Loss of Secondary Heat Sink) has been entered and feed and bleed of the NC system was initiated
Shortly after opening the PORV5, the Turbine Driven CA pump is returned to service and a source of feedwater is available
CETsarestable
All S/G WR levels are indicating 0%
Containment pressure is 3.5 PSIG

1. Based on the conditions described above which ONE (1) of the following describes the criteria for restoration of CA flow?

2. What is the basis for the restoration of flow criteria?

A. 1. Restore cooling to ALL intact S/Gs at a rate not to exceed 100 GPM

2. To minimize additional NC cooldown causing thermal stress to the reactor vessel B. 1. Restore cooling to ALL intact S/Gs at a rate not to exceed 100 GPM

2. To minimize the thermal stress on the S/G to prevent failure of S/G corn pon ents C. 1. Restore cooling to ONE intact S/G at a rate not to exceed 100 GPM

2. To minimize additional NC cooldown causing thermal stress to the reactor vessel D. 1. Restore cooling to ONE intact S/G at a rate not to exceed 100 GPM

2. To minimize the thermal stress on the S/G to prevent failure of S/G components Tuesday, July 13, 2010 Page 133 of 294

___________ _____ ___ ___________ __________ _ __________ ___________ _____ ___ __ ______ ___________ _______ ___________ __ FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 47 General Discussion In the scenario given, a loss of feedwater/heat sink had occurred. FRP H-I has been implemented and feed and bleed was established. When the capability to feed from AFW is restored the procedure contains a continuous action statement in the RNO for Step 7 e to return to step 7.h. is CA is restored and Step 35 has been performed which it would have since feed and bleed has been established. With all SIG <17% WR level (all dry), H-l directs that flow be established to one intact S/G at less than or equal to 100 GPM. There is a note prior to this step concerning the risk of thermal shock to the S/G and also a caution in FR H-5 (Response to SIG Low Level) which reads Initiating feed flow to a dry SIG causes thermal stresses and raises the risk of S/G failure, especially on the SIG shell. The risk is greatest at [her SIG temperatures. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: First p is correct for flow rate but only one S/G will be fed; if CETs were increasing then feeding all of intact SIGs would be correct and therefore plausible. Second part is plausible because overcooling the NC system is stated in H-i as a concern in multiple notes and cautions concerning initiating oa dry generator. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: First part is correct for flow rate but only one S/G will be fed; if CETs were increasing then feeding all of intact S/Gs would be correct and therefore plausible. nd part regarding the basis is correct and therefore plausible. Answer C Discussion COCT: See explanation above. AUSIBLE: First part is correct. Second part is plausible because overcooling the NC system is stated in H-l as a concern in multiple notes and cautions concerning initiating ed to a dry generator. - Answer D Discuss!on CORRECT: See explanation above. - Basis for_meeting the KA fi. is matched because a loss of feedwater has occurred and the question is testing knowledge related to how many SIGs will initially be fed (operational implication) and what concern is being addressed_by this strategy (effects of feedwater introduction on dry S/G). Basis for Hi Cog This is a hi cog question because it involves a level of analysis of given situation. This involves a multi part mental process where the applicant must evaluate the indications given and determine its meaning related to the scenario given and determine a course of action and the basis for that action. Basis for SRO only Job Level Cognitive Level QuestionType I Question Source RO Comprehension NEW

                                         .1 rDevelopment References               -

Student References Provided From FRP H-i Page 6 of 93

  -MC-EP-FRH Obj: 4                       -

APEO54 AKI.02 Loss of Main Feedwater (MFW)

Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater (MFW): (CFR 41.8 / 41.10 / 45.3) Effects of feedwater introduction on dry S/G Tuesday, July 13, 2010 Page 134 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 47 2547 01-9 Comments: Remarks/Status I______________________________________________________ _ 1-9 Comments: No comment Resolution / Comments: N/A Tuesday, July 13, 2010 Page 135 of 294

Question 47

References:

OP-MC-EP-FRH Obj: 4 OBJECTIVES s NNLLL E OBJECTIVE L L P P 0 OORSR Q RO0 I Explain the purpose of each procedure in the FR-H series. X X EPFRHOO1 2 Discuss the entry and exit guidance for each procedure in the X X FR-H series. EPFRHOO2 3 Discuss the mitigating strategy (major actions) of each X X X procedure in the FR-H series. EPFRHOO3 4 Discuss the basis for any note, caution or step for each X X X procedure in the FR-H series. EPFRHOO4 5 Given the Foldout page, discuss the actions included and the X X X basis for these actions. EPFRHOO5 6 Given the appropriate procedure, evaluate a given scenario X X X describing accident events and plant conditions to determine any required action and its basis. EPFRHOO6 7 Discuss the time critical task(s) associated with the FR-H X X X series procedures including the time requirements and the basis for these requirements. EPFRHOO7

From FRP H-I Page 6 of 93 MNS RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO. EP/l /A! 5000IFR-H.l 6 of 93 Rev. 13 UNIT 1 ADTION/EZPECIED EsroNsE ES;tNsE NOT o3TA:NE:

7. (Continued)

h. Check any 510 W/R level LESS THAN

                                             -                        h. Perform the following:

12% :17% ACC).

1) Throttle open CA control valves to establish CA flow to S/Os.

_2) GO TO Step37. NOTE

It may be preferable to feed 1 B or 1C 510 first, to maintain steam supply for TD CA pump Selecting S.G with highest level will reduce risk of thermal shock to 510 vhen reestablishing feed flow I. Check core exit TiCs STABLE OR

                                        -                            L    erforni the followina:

GOING DOWN.

1) Throttle open CA control valve to one 510 to establish flow rate required to lower core exit TiCs.

2) jf core exit TCs continue to go up, THEN throttle open CA control valve to feed another S/G as required to lower core exit TiCs.

3) GO TO Step 7.m.

j. Slowly throttle open CA control valve to one 510 to establish feed flow less than or equal to 100 GPM.

Ic Maintain teed flow rate less than or equal to 100 3PM until St WRIevel is greater than 12% (17% .4CC). I. WHEN 510 WIR level is greater than 12% 17% ACC). THEN feed flow may he raised greater than 100 GPM. nt Check 810 ?diR levels on intact SiGs ro. GO TO Step 7.0. with feed flow isolated ANY

GREATER THAN 12% (17% ACC).

n. Slowly establish flow to any available intact S/G with level greater than 12%

(17% ACC).

From FRP H-5 Page 3 of 4 MNS RESPONSE TO STEAM GENERATOR LOW LEVEL PAGE NO. ER:1IN5QOCJFRH.5 3 of 4 nrri ACTZON/EXECED RESPONSE RESCNSE NOT OBTAInED

4. Check affected SiC(s) INTACT:

                                 -                   IF affected SiC pressure going down in an uncontrolled manner OR is
       . Affected SIG pressure  - STABLE OR       depressuthed, THEN:

GOING UP

IF affected S!G(s) previously identified as

Affected SIG - PRESSURIZED. fauted, IHLK RETURN EQ. procedure in effect.
IF affected S?G:s:t has not been isolated.

THEN GO TO EPI1!A/6000/E-2 (Faulted Steam Generator Isolationi.

5. Check CA flow to affected SIG(s) - Perform the following:

GREATER THAN 25 GPM.

a. IF affected SIG:sj W:R ,evel s greater than 12% (17% ACC). THEN establish CA flow as necessary to refill affected CAUTION Initiating feed flow to a dry SIC causes thermal stresses and raises the risk of SIG failure, especiaLly on the SiC shell, The risk is greatest at higher S1G temperatures.

b. IF affected S/Gcs, AR level is less than 12% (17% ACC:. THEN:

Leave feedwater isolated to affectea SiGs;. 2: Contact station management to evaLiate refilling affected S!Gsi as pan of long-term plant recoverj. 3: RETURN TO procedure and step in effect.

6. Continue raising affected SIG(s) NIR level to greater than 11% (32% ACC).

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 48 ( EPEO55 EA2.O1 Loss of Offsite and Onsite Power (Station Blackout)

bility to determine or interpret the following as they apply to a Station Blackout: (CFR 43.5 I 45.13) xisting valve positioning on a loss of instrument air system Given the following plant conditions:

Due to a fault at the switch yard, the site has experienced a LOOP
Unit 1 subsequently lost both D/Gs
Due to a rupture of the Diesel VI compressor discharge piping, VI header pressure is indicating 0 PSIG
The crew is performing ECA 0.0 (Loss of All AC Power)
Prior to this event, Unit 1 was at 100% RTP with normal LID in service and flow being controlled with 1 NV-459 (UI Variable LID Orifice Outlet Flow Cntrl)
The Crew is performing Step 6 of ECA 0.0 Check NC System ISOLATED -

Assuming no manual operator action has been taken associated with these components, which ONE (1) of the following correctly lists the expected As Found positions for the valves listed below? o 1NV-35A (Variable L/D Orifice Outlet Cont lsol) o 1NV-1A (NC L/D Isol To Regen Hx) A. 1 NV-35A CLOSED

1NV-1A -CLOSED B. INV-35A-OPEN 1NV-1A -CLOSED C. 1NV-35A-CLOSED 1NV-1A -OPEN D. 1NV-35A-OPEN 1NV-1A -OPEN Tuesday, July 13, 2010 Page 136 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 48 2548 neral Discussion With conditions given, there has been a loss of all AC (Station Blackout) along with a loss of VI. For this question had to include a VI header rupture in addition to Blackout. Since M1JS has Diesel VI compressors it is not plausible to lose VI system pressure soley because a Blackout has occurred. The applicant is asked to determine the expected postions of 2 valves in the letdown section of the CVCS system. Both of these valves would be required to be checked per ECA 0.0 Step 6. Even though both are powered from EVDA which is a bus which would remain energized with the conditions given, both of these valves would be closed. Both are supplied from VI which has been lost and would result in both valves failing closed. Answer A Discussion CORRECT: See explanation above. Answer B Discussion - -____________ INCORRECT: See explanation above. PLAUSIBLE: 1NV-35 would remain energized via vital batteries and would not have been manually closed. The applicant could conclude that it would therefore remain open. Position is correct for 1NV-1A. Answer C Discussion - -________________________________________________________________ INCORRECT: See explanation above. PLAUSIBLE: Position is correct for INV-35A it would be closed. The applicant could conclude that INV-1A would remain open because it would have been open prior to the event and as stated in the stem, no manual action was taken to close it. Like 1NV-35A it would have remain energized during the event. -______________________________________________________ Answer D Discussion -_____________________________________ INCORRECT: See explanation above. ( AUSIBLE: 1NV-35 would remain energized via vital batteries and would not have been manually closed. The applicant could conclude that vould therefore remain open. The applicant could believe that 1NV-1A would remain open because it would have been open prior to the event dnd as stated in the stem, no manual action was taken to close it. Like 1NV-35A it would have remain energized during the event. Basis for meeting the KA K/A is matched because the question has placed the plant is a situation where both a station blackout has occurred and a loss of lAS. The applicant is then asked to determine the existing valve positions of two NV valves which have not been manually manipulated. 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 existing valve positions for the two valves in question. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References ient References Provided OP-MC-PS-NV Page 23 (Rev 58) From OP-MC-PS-NV Page 55 (Rev 58) ECA 0.0 Page 4 of 174 (Rev 26) AP-22 (Loss of VI) Page 107 of 121 (Rev 28) AP-22 (Loss of VI) Page 107 of 121 (Rev 28) OP-MC-PS-NV Obj: 7 E055 EA2.01 Loss of Offsite and Onsite Power (Station Blackout)

ility to determine or interpret the following as they apply to a Station Blackout: (CFR 43.5/ 45.13) Existing valve positioning on a loss of instrument air system Tuesday, July 13, 2010 Page 137 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 48 2548 401-9 Comments: Remarks!Status 401-9 Comments: No comment. Resolution / Comments: N/A Tuesday, July 13, 2010 Page 138 of 294

Question 48

References:

OP-MC-PS-NV Obj: 7 NNLLL E OBJECTIVE L L P P 0 Q OORSR R00 5 Explain the basic operation of the NV System for the x x x x following:

   . Normal L.D. Purification
   . Seal Injection Flow
   . Chemical Addition
   . Charging
   . Centrifugal Charging Pumps
   . All Modes of Makeup
   . PD Pump Control
   . Safeguards Actuation
   . Charging/Letdown Flow Balance
   . Excess Letdown
   . Emergency Boration
   . Pressurizer Spray 6   Describe the various system parameters indicated in the            X X X Control Room associated with the NV System in ALL modes of operation.

7 List the fail position of NV valves on loss of power or air. X X X 8 Describe the as-built configuration of the VCT level X X X X instrumentation. 9 Using fundamental instrumentation knowledge and given X X X X specific reference and variable leg configurations for the Volume Control Tank, predict the effect on indicated versus actual level for various failures.

From OP-MC-PS-NV Page 23 (Rev 58) 2.3 Letdown Orifice I Letdown Throttle Valves Objective #4 The letdown orifice I letdown throttle valves are designed to reduce the NC system pressure by 1900 psig and to control the letdown flow. The orifice reduces flow to 45 gpm and is isolated by NV-457A. One letdown throttle valve, NV-454, is manually set at 75 gpm via the Valve Checklist OP. It is isolated by NV-458A. In addition, there is a flow control valve, NV-459, that is controlled by a manual loader on the control board or on the Auxiliary Shutdown Panel (ASP). The flow control valve, NV-459, allows the operator to control flow when heating up the letdown path to avoid thermal shock (water hammer, etc.) and provides for increased letdown flow during low-pressure operation. NV-459 is also the preferred flow path during normal operation. The letdown orifice isolation valves (NV 457A, NV-458A, NV-35A) are each controlled by a three position switch (Open-Automatic-Close) from the Control Room or the ASP. The ASP has a Remote-Local switch. They function as containment isolation valves in addition to providing a means to isolate the orifice I letdown throttle valves. NV-457A, NV-458A and NV-35A have the following interlocks:

Auto close on Low Pzr Level (17%)
Auto close if NV-iA or NV-2A closes
Auto close on Phase A isolation (Si).

In a Loss of Letdown event (AP-12) with the orifice isolation valves going closed, it may become necessary to locally pressurize the letdown header from the charging header, in order to prevent water hammer. NV-i 06 (a manual valve in the pipe chase) will allow the repressurization of the letdown line from the charging header. NV-6 serves as over-pressure relief protection for the letdown piping downstream of the letdown isolation valves. Relief setpoint is 600 psig and it relieves to the PRT. 2.4 NV-7B (LID Containment Isolation Valve) NV-7B closes on a Phase A Containment Isolation signal (Si) and is normally controlled from the Control Room. 2.5 NV-121 (LID from ND System) NV-121 allows letdown from the ND System for NC system cleanup when the differential pressure across the orifices is too small. Also it is used to initially pressurize the ND system when placing it in service during unit shutdown.

From OP-MC-PS-NV Page 55 (Rev 58) NV-i 047A Recirc Valve NV-i 047A has Open-Close pushbuttons. The valve will

close 2 minutes after the PD Pump starts. PDP Control Board M/A Station In Manual, the raise/lower pushbuttons are used to

control PD Pump speed. The PD Pump is always operated in Manual, since IAE does not maintain the Auto portion of the PD Pump Speed Controller. In the Auto mode, the Pressurizer Level Master controller input controls PD pump speed. However, Auto is not used. A Suction Dampener was installed to reduce vibration of the PD Pump. A local On/Off switch, with associated indicating lights, controls the Suction Dampener heater. NV Lube Oil Pumps The CCP auxiliary lube oil pumps are controlled from the Main

Control Board, MC-1O, with Auto-Man-Start-Stop pushbuttons. In Auto, the lube oil pump will start if the CCP is running and lube oil pressure is < 8 psig. NC Letdown Isolation Valves (NV-IA, NV-2A) Each valve is controlled from the Main

Control Board, MC-1O, with a 3-position switch, Open-Auto-Close, with spring return to Auto. They may also be controlled from the ASP. There is a Remote-Local switch at the ASP to determine control. These valves close on Low Pressurizer Level of 17%. They will Fail Closed on loss of power. NV-lA closes when control is transferred to the SSF. Letdown Orifice Isolation Valves (NV-457A, NV-458A, NV-35A) Each valve is

controlled from the Main Control Board, MC-1O, with a 3-position rotary switch, Open-Auto-Close with spring return to Auto. The valves may also be controlled from the ASP. There is a Remote-Local switch at the ASP to determine control. These valves will Auto Close on the following: 1) an S signal, 2) Low Pressurizer Level (17%), and 3) if NV-iA or NV-2A close. The Letdown Orifice Isolation Valves must have a Full-Open indication prior to releasing the switch Open position to prevent reclosure of the associated valve. NV-459 The Letdown Flow Control Valve is controlled from the Main Control Board,

MC-i 0, by a Manual Loader. The valve may also be controlled from the ASP. This valve fails closed. The valve will close on a loss of KXA for Ui and on a loss of KXB on U2. NV-7B The Letdown Containment Outside Isolation Valve is controlled from the Main

Control Board, MC-10, by Open-Close pushbuttons. This valve will Auto-Close on an ST signal. NV-124 The Letdown Pressure Control Valve is controlled by a Manual-Auto Station

Pressure Controller on MC-i 0. In Auto, the controller receives input from a pressure transmitter downstream of the Letdown Heat Exchanger. In Manual, the operator positions NV-i 24 with the Open/Close pushbuttons on the M/A station. The valve fails open. NV-i 27A The Letdown Heat Exchanger Outlet 3-Way Temperature Control Valve is

controlled by a 3-position rotary switch, VCT-Normal-Demin, on MC-10 with spring return to Normal. The valve automatically shifts to the VCT if a high temperature occurs in the NV letdown line. The valve fails to the VCT position on a loss of air.

From ECA 0.0 Page 4 of 174 (Rev 26) MNS LOSS OF ALL AC POvER PAGE NC. EP!IIAI50001ECA-O.O 4 of 174 Rev. 26 UNIT 1 ACTION/EXECIED RESEONSE E3EtNSE NOT O3METh

6. Check NC System ISOLATED:

a. Check the following letdown orifice a. CLOSE valvet.

isolation valves CLOSED:

1) 1NV-456A (75 GPM LID Orifice Outlet Cont 1501).

2: 1NV-457A $5 GPM LID Orfice Outlet Cont lsol).

3) 1NV-35A (Variable LID Orifice Outle: Cont Isol).

b. CLOSE the following valves:

1) 1NV-1A (NC LID Isol To Regen Hx).

2: 1NV-2A :NC LID Isol To Regen Hx:.

c. Check P:r PCRVs-CLOSED. c. IFP:rpressure less than 2315 P51G.

THEN CLOSE all P:r PORVs.

d. Check the following excess letdown d. CLOSE valve(s).

isolation valves CLOSED:

                                -
            . 1NV-24B (C NC Loop To E:xs LID H..:

Isol)

            . 1 NV-25B (C NC Loop To Exs LID Hx Isol :.

e. Check 1NV-121 (Ui ND Letdown e. CLOSE valve.

Control) CLOSED.

From AP-22 (Loss of VI) Page 107 of 121 (Rev 28) MNS LOSS OF VI PAGE NO. AP,11A15500122 121 Enclosure 12- Page 3 of 6 Valve Failure Mode on Loss of Air Re28 UNIT 1

8. MV valves:

a. The following NV valves fail open:

          . 1NV-16A (NV Supply To D NC Loop 1501)
          . 1NV-13B (NV Supply T0ANC Loop lsol(
          . 1NV-34A (A NC Pump Seal Return Iso))

             . 1NV-SOB :B NC Pump Seal Return Isol)
          . 1NV-66A (C NC Pump Seal Return Isol)
          . INV-82B CD NC Pump Seal Return Isol)
          . INV-124 (Letdown Pressure Control
          . 1NV-238 Charging Line Flow Controli

1NV-241 (Ui Seal Water lnj Flow Controb

             . 1NV-267A (Boric Acid To Blender Control:.

b. The following NV valves faU to the VCT position:

             . INV-27B (Excess liD Hx orn 3-Way Cntrli

             . 1NV-127A (LID Hx Outlet 3-/lay Temp Cntrl)

             . 1NV-137A (NC Filters OtIt 3-Way Cntrl:.

c. The following NV valves fail closed:

          . 1NV-1A:NCLDlsolToRegenb:.:C
          . INV-2A (NC LID Isol To Reqen Hx)
          . 1NV-21A (NV Spray To PZR Isol) 1NV-24B (C NC Loop ToExs LID Hx Isol:
          . 1NV-25B (C NC Loop To Exs LID Hx Isol:

             . 1 NV-26B (Ui Excess L?D Hx Outlet Cntrl:

             . INV-35A (Variable LID Orifice Outlet Cont Isol)

             . 1NV-39A (A NC Pump Standpipe Fill)
          . 1NV-55B (B NC Pump Standpipe Fill)
          . INV-71A (C NC Pump Standpipe Fill)
          . 1 NV-67B CD NC Pump Standpipe Fill:,
          . 1NV-92A (NC Pumps Seal Byp Return Hdr Isol)

             . 1 NV-i 21: Ui ND Letdown Control:

             . 1NV-i67A (VCT Vent To WG lsol(
          . 1NV-171A (BAblenderToVCTlnlez)

e 1NV-i7SA (BA Blender to VCT Outlet) 1NV-457A (45 6PM LID Orifice Outlet Cont Isol)

         . 1NV-$58A (75 GPM liD Orifice Cutlet Cont Isol)

1NV-459 (Ui Variable LD Orifice Cutlet Flow Cntrl)

         . iNV-840A(U1 ND To Pzr Aux Spray Comrol:.

From AP-22 (Loss of VI) Page 107 of 121 (Rev 28) MNS LOSS OF VITAL OR AUX CONTROL POWER PAGE N0 AP/IIAJ5SQO:i5 EnclosureS- Page 4 of 7 1EVDA Load List R UNIT 1

10. NV System:

The following valves fail closed:

         . 1NV-1A (NC Ut Isol To Regen Hx)

            . 1NV-2A (NC Ut Isol To Regen Hx)

1NV-457A (45 GPM LID Orifice Outlet Cont sal)

            . INV-458A (75 GPM L/D Orifice Outlet Cont Isol)

INV-3sAçariahle LiD Orifice Outlet Cont Isoli

            . I NV-171A (BA Blender To VCT Inlet:

INV-i75A (BA Blender to VCT Outlet)

I NV-252A (Rx NW Water To Blender Contrail

            . 1NV-167A kVCT Vent To         G sob

            . INV-39A (A NC Pump Standpipe Fill:

            . 1NV-71A (C NC Pump Standpipe Fill:

1NV-21A (NV Spray To PZR Isol)
1NV-92A (NC Pumps Seal Byp Return Hdr Isol)

            . 1NV-840A (Ui ND To P:r Aux Spray Controli.

The following valves fail open:

         . 1NV-1SA(NV Supply ToD NC Loop Isol)

            . INV-267A (Boric Acid To Blender Control:

            . 1NV-2$AA NC Pump Seal Return soIl

            . INV-66A (C NC Pump Seal Return soIl.

The following valves fail to position:
1NV-127A :LD H;.:. Ouet 3-Way Temp Cntrl)
1 NV-i 37A (NC Filters CU: 3-Way Cntrl)

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2010 MNS SRO NRC Examination QUESTION 49 APEO56 AA2.50 Loss of Offsite Power

( bili to determine and interpret the following as they apply to the Loss of Offsite Power: (CFR: 43.5 / 45.13) that load and VAR limits, alarm setpoints, frequency and voltage limits for ED/Gs are not being exceeded Given the following conditions on Unit 1:

A loss of off-site power has occurred
1A and I B DG5 have started and loaded normally Based on the following loading profile for 1A DG:

DG 1A LOAD (KW) 5000 4800 4600 EEEEEf:E 4400 4200 4000 3800 3600 3400 4:00 5:00 6:00 7:00 8:00 9:00 10:00 DG 1A LOAD (KW)

1. The maximum design load limit for CONTINUOUS operation was FIRST exceeded at (1)

2. The maximum design load limit for operation in an OVERLOAD condition was FIRST exceeded at (2)

Which ONE (1) of the following completes the statements above? A. 1. 0430

2. 0600 B. 1. 0600

2. 0745 C. 1. 0430

2. 0745 D. 1. 0600

2. 0830 Tuesday, July 13, 2010 Page 139 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 49

,eneral Discussion The maximum continuous load for an DO is 4000 KW. The DOs may be operated at up to 4400 KW for two hours in a 24 hour period.

The first time that the DO exceeded the continuous load limit of 4000KW was at 6:00. The DO exceeded the design overload limit of 4400KW at 7:45. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because the DG surveillance requires the DO to be tested at a minimum of 3800 KW to meet operability requirements. Part 2 is plausible because 4000KW is the maximum continuous load limit. Answer B Discussion CORRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because the DO surveillance requires the DO to be tested at a minimum of 3800 KW to meet operability requirements. Part 2 is correct. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: Part I is correct. rt2 is plausible if the applicant concludes that the maximum design overload limit is 4500 KW instead of 4400 KW. asis for meeting the KA The KA is matched because a Loss of Offsite Power has occurred and the applicant must know the load requirements for the DGs and determine those limits have been exceeded. Basis for Hi Cog This is a higher cognitive level because it required multiple mental steps to arrive at the correct answer. First, the applicant must recall from memory the limit for operating an EDO in an over-load condition (greater than 4000 KW but less than 4400 KW for 2 hours /24 hours). The applicant must then analyze the load profile for the EDO and determine when the DO exceeded the continuous load limit and maximum design overload limit. Basis for SRO only Job Level__ritive Level QLiestionType Question Source RO Comprehension NEW Development References Student References Provided Lesson Plan OP-MC-DO-DG Section 2.1 APEO56 AA2.50 Loss of Offsite Power

Ability to determine and interpret the following as they apply to the Loss of Offsite Power: (CFR: 43.5 / 45.13) That load and VAR limits, alarm setpoints, frequency and voltage limits for ED/Os are not being exceeded 9 Comments: R RemarkslStatus 401-9 Comments: No comment. Tuesday, July 13, 2010 Page 140 of 294

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2010 MNS SRO NRC Examination QUESTION 49 Resolution I Comments: Tuesday, July 13, 2010 Page 141 of 294

Question 49

References:

From Lesson Plan OP-MC-DG-DG Section 2.1:

1.0 INTRODUCTION

11 Purpose Objective # 1 7 The purpose of the Diesel Generators is to provide standby AC power to the equipment required to safely shut down the reactor in the event of a loss of normal power source. The Diesel Generators will also supply power to the safeguards equipment as required during a major accident coincident with a loss of normal power source. 1.2 General Description Objective # 2 At McGuire, two onsite diesels per unit are provided to respond to basically three major accident situations:

1. A Loss of Coolant Accident

2. A Blackout (loss of voltage to safeguards bus)

3. A combination Loss of Coolant Accident and a Blackout.

During a LOCA both diesels start and run but if normal power is available they will not close in on the bus. During a Blackout both diesels will start, run, close in on the bus, and remain that way until the problem has been resolved. During a Blackout followed by a LOCA the diesel generator will trip all non-LOCA loads and pick up all the LOCA loads not sequenced on by the Blackout Sequencer. If there is a LOCA followed by a Blackout the diesel will pick up the LOCA loads that were being supplied prior to the Blackout. Each diesel also has Local and Remote Manual loading capability. NOTE: Both unit diesels have identical controls and instrumentation systems. 2.0 FUNCTIONAL DESCRIPTION 2.1 Design Generator Engine 4160 VAC, 3 Phase, 60 HZ 16 cylinders 4000 kW @ 0.8 pf Continuous power 514 RPM rated speed 4400 kW 0.8 pf for 2 hours/24 hours Over- 28 psig minimum operating lube oil pressure load Capability Phase Differential (87G), and Overcurrent 13.5 Bore! 16.5 Stroke Protection (51V) 125 VDC, Field Flash @ 40% speed Excitation 5575 BHP

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 50 APEO58 2.1.27 Loss of DC Power

PE058 GENERIC now1edge of system purpose and/or function. (CFR: 41.7) The Kirk Key interlocks located on the Vital Battery Charger Connection Boxes (ECB-1 thru 4) associated with EVCA, EVCB, EVCC and EVCD prevent Which ONE (1) of the following completes the statement above? A. supplying A Train Busses from the B Train Source B. tying a Unit 1 power source to a Unit 2 power source C. energizing more than one battery from the Standby Charger D. supplying two 1 25v DC Distribution Centers from the Standby Charger Tuesday, July 13, 2010 Page 142 of 294

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2010 MNS SRO NRC Examination QUESTION 50 255f

 .eneral Discussion The charger connection box breakers for EVCA, EVCB, EVCC, and EVCD are Kirk-Key Interlocked to allow only one breaker to be closed at a This prevents tying a Unit I power source to a Unit 2 power source.

Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because this is a function of the Kirk-Key interlock for the standby battery charger (EVCS). Answer B Discussion 5RREcT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because this is a function of the Kirk-key interlock associated with the standby charger (EVCS). However, it is NOT a function of the Kirk-Keys for the Vital Battery Charger connection boxes. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because this is a possible alignment with the standby charger. Basis for meeting the KA On a loss of DC power due to a fault, the Kirk Keys limit the impact of the loss of DC power by preventing cascading losses of DC equipment. Therefore,_the K/A as it relates to knowledge of the purpose or function and a Loss of DC Power is met. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source L_ RO Memory BANK M1JS Exam Bank #ELEPLOI5 Development References Student References Provided Learning Objective:

1) EL-EPL #9

References:

I) Lesson Plan OP-MC-EL-EPL Section 2.1 APEO58 2.127 Loss of DC Power

APEO58 GENERIC Knowledge of system purpose and/or function. (CFR: 41.7) 401-9 Comments: Remarks!Status -_________ 40 1-9 Comments: No comment. Resolution / Comments: Tuesday, July 13, 2010 Page 143 of 294

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2010 MNS SRO NRC Examination QUESTION 50 Tuesday, July 13, 2010 Page 144 of 294

Question 50

References:

From Lesson Plan OP-MC-EP-EPL Section 21: Since EVCS (spare battery charger) is identical to the other chargers it can be used to replace a normal charger (as necessary) by closing the appropriate key interlocked circuit breakers. Objective#8 I The load demands normally placed on each unit battery charger will consist of its respective DC distribution center loads, as well as, the loads of the associated DC panelboards while still providing a floating charge (132 +/- 1 volt) on its respective battery. Each charger receives power from one of two redundant 600 VAC Essential Auxiliary Power System Motor Control Centers (one power supply from a Unit 1 MCC and the other from a Unit 2 MCC). The chargers are manually connected to either one of these two power supplies through their respective charger connection box. 1(2) EMXA are the MCCs feeding the connection boxes for EVCA and EVCC, while the MCCs feeding the connection boxes for EVCB and EVCD are 1(2) EMXB. Objective # 9 I The charger connection box breakers for EVCA, EVCB, EVCC, and EVCD are Kirk-Key Interlocked to allow only one breaker to be closed at a time. This prevents tying a Unit I power source to a Unit 2 power source. Charger startup involves closing the DC output breaker to the Distribution Center then the charger AC input breaker. The control board operator will then start the battery charger by depressing the start push-button, located on 1MC-8 in the Control Room, which closes a set of m contacts, located at the 600 V MCC, and provides AC power to the battery charger via the charger connection box. Then the Charger DC output breaker is closed connecting the charger to the DC loads. Charger shutdown requires the control board operator to depress the stop push-button, located on 1MC-8 in the Control Room, followed by opening of the charger DC output, AC input breaker and then the DC output breaker to the Distribution Center. Objective # 10 & 11 The standby charger (EVCS) is used when one of the normal battery chargers is unavailable for service (standby mode) or during an equalizing charge to one of the batteries. The two feeder breakers, located at EVDS (distribution center for battery charger EVCS), provide proper alignment of the standby charger during its operation (standby mode or equalizing charge mode). The standby charger can supply the A Train Distribution Centers (EVDA or EVDC) or the B Train Distribution Centers (EVDB or EVDD). Kirk-Key Interlocks, provided with all of the associated breakers, ensure that only one train of distribution centers can be supplied, from EVDS, at a time.

In the standby mode of operation EVCS will replace the out-of-service battery charger. During this mode of operation the out-of-service battery charger is disconnected from its distribution center with the spare charger connected to the distribution center through one of the distribution center (EVDS) breakers, discussed above. In addition, the tie breaker to the distribution center with the out-of-service battery charger must be closed. During the equalizing charge mode the normal battery charger is disconnected from its distribution center and will be aligned in parallel with its respective battery. The normal battery charger will be placed in Equalize mode of operation. Battery charger (EVCS) will then supply the distribution center with the tie breakers closed (cross-tied with its sister channel). This same alignment is utilized during normal charger maintenance and battery discharge testing. During these operations, the normal charger remains in the Float mode. Objective # I As discussed above, the breakers, associated with standby battery charger EVCS are Kirk-Key Interlocked. Referencing Training Drawing 7.1, Composite Vital I/C Drawing, may help in your understanding of the interlocks described below:

The breakers at distribution center EVDS are Kirk Key Interlocked with each other and their respective connection box (ECB5) such that:

1) The A Train feeder breaker from 1 EMXH in ECB5 cannot be closed unless the A Train supply breaker for EVDA or EVDC (located at distribution center EVDS) is closed. This prevents the A Train source from supplying the B Train buses.

2) The B Train feeder breaker from 2EMXH in ECB5 cannot be closed unless the B Train supply breaker for EVDB or EVDD (located at distribution center EVDS) is closed. This prevents the B Train source from supplying the A Train buses.

3) Only one breaker from EVDS can be closed at a time. This prevents the standby charger from supplying both A Train and B Train buses.

In addition, the supply breakers to ECB5 (Connection Box) from IEMXH and 2 EMXH are Kirk-Key Interlocked to prevent closure of both breakers at the same time. This interlock scheme in conjunction with I & 2 above prevent cross connection of A & B Train AC sources and minimizes mutual exposure of the two trains.

2.2 125 VDC Vital Instrumentation and Control Power System Batteries Both units (Unit 1 and 2) are provided with only four 125 VDC Vital Instrumentation and Control Power System batteries. Each battery consists of 60 total cells; with each cell packaged in a clear plastic, non-combustible, shock-absorbing container with the appropriate covers, racks, and accessories. The battery is connected to its respective DC distribution center, in parallel with its respective battery charger, and located in an individual and physically separate room within the main battery room.

Question 50 Parent Question: ELEPLOI5 1 Pt The Kirk Key interlocks associated with the Vital Battery Charger Connection Boxes (EVCA, EVCB, EVCC, EVCD) prevent: A. Paralleling the Standby battery charger with a normal battery charger. B. Tying a Unit 1 power source to a Unit 2 power source. C. Energizing more than one battery charger from the same power source. D. Paralleling two batteries with one normal battery charger. Answer 146 B

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 51 APEO62 AK3.04 Loss of Nuclear Service Water

(Nnowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water: (CFR 41.4, 41.8 /45.7) ffect on the nuclear service water discharge flow header of a loss of CCW Given the following conditions on Unit 1:

The unit is at 100% RTP
Train swap is in progress and currently both trains of KC and RN have been placed in service.
The 1A RN pump TRIPPED
A Unit 2 electrical fault causes a B/C associated with 2ETA.

Based on the conditions described above and assuming no operator action, which ONE (1) of the following describes the effect of this event on Unit 1? A. Cooling flow would be lost to the 1A KC HX due to Unit 1 RN Train separation. The lB RN Train suction and discharge alignment would be unaffected. B. Cooling flow would be lost to the IA KC HX due to Unit 1 RN Train separation. The lB RN Train suction and discharge would realign to the SNSWP. C. The 1 B RN Pump would continue to supply cooling for the 1A KC HX because ( the Unit 1 RN Train cross connect valves remain open. The lB RN Train suction and discharge would realign to the SNSWP. D. The lB RN Pump would continue to supply cooling for the 1A KC HX because the Unit 1 RN Train cross connect valves remain open. The lB RN Train suction and discharge alignment would be unaffected. Tuesday, July 13, 2010 Page 145 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 51 255I

 .eneral Discussion In the stem, the applicant is presented with a situation where both trains of RN and KC were placed in service on U-I. The 1A RN pump has tripped but initially the lB RN will provided cooling to both trains of KC via normally open RN train cross connect valves. A BIO then occurs on U-i. This would result in the A Train of RN on both units aligning to LU (normally aligned there so no change) and the B Train of RN on both units realigning to the SNSWP. The signal would also result in train separation on both units (1 and 2 RN-41A would close) resulting in a loss of cooling to the 1A KC HX because the 1A RN pump is unavailable.

Answer A Discussion - INCORRECT: See explanation above. PLAUSIBLE: First part of the answer is correct with the correct reason and therefore plausible. The second part is plausible if the applicant thinks the U-2 signal only affects the U-2RN alignment. This is not a unreasonable assumption. Answer B Discussion -- CORRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: First part of the answer is plausible if the candidate fails to remember that the U-2 signal affects train separation on both Units. The second part of this distracter is correct and therefore plausible. -- Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: First part of the answer is plausible if the candidate fails to remember that the U-2 signal affects train separation on both Units. The second part is plausible if the applicant thinks the U-2 signal only affects the U-2 RN alignment. This is not a unreasonable assumption. asis for meeting the KA ie K/A is matched because the applicant is presented with a scenario where, because of an unusual alignment and the introduction of a B/O signal, RN is lost to the U- 1 KC HX. The applicant must demonstrate an understanding of the reason for the loss. Also the loss of cooling affects the CCW discharge flow header. Basis for Hi Cog The question is Hi cog because the applicant must analyze a given scenario and predict an outcome. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided OP-MC-PSS-RN Page 49 (Rev 43) - Lesson Plan OP-MC-PSS-RN Page 95 (Rev 43) APEO62 AK3.04 Loss of Nuclear Service Water

Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water: (CFR 41.4, 41.8 / 45.7) Effect on the nuclear service water discharge flow header of a loss of CCW 401-9 Comments: RemarkslStatus Chief Examiner approved use of reverse logic on this KA to be able to write an operationally valid question (i.e. the effect of a loss of RN on CCW) 02/19/10 401-9 Comments: Tuesday, July 13, 2010 Page 146 of 294

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2010 MNS SRO NRC Examination QUESTION 51 Eicond part of A needs the word train after RN. Resolution / Comments: Corrected answer A per Lead Examiners comment. See attached file for revised copy. Tuesday, July 13, 2010 Page 147 of 294

Question 51

References:

OP-MC-PSS-RN Obj: 8 OBJECTIVES NNLLL OBJECTIVE L L P P 0 OORSR ROO 8 Describe the RN System Flow path ( suction source, essential and non-essential header alignment and discharge point) for the following:

        . Normal operation                                         X X X X
       . Operation following a Blackout                           X X X X X
       . Operation following a Safety Injection                   X X X X X 9  Explain the reason for taking a suction on the low level intake. X X X X 10  Concerning the RN essential and non-essential headers:
       . List the loads supplied by each header                   x x x x
       . Identify which loads are automatically supplied on a     x x x x x Blackout, Safety injection and/or Phase B.

11 Explain the reason for NOT isolating the auxiliary building X X X X X non-essential header on a Blackout signal. 12 Describe the operation including any interlocks for the X X X X X following va(ves:

       . RN42A ( AB Non Ess Supply Isol)
       . RN171B (B DIG Supply 1501)
       . 1RN1 ( Low Level Intake Isolation)
       . Engineering Safeguards Modulating Control Valves and Reset Circuitry 13  Describe the operational concerns when cycling RN valves             X X X that are shared between Unit 1 and Unit 2.

14 Given a parameter associated with the RN system, describe X X X X the indications for that parameter. 15 Given a Limit and Precaution associated with the RN System, X X X X X discuss its basis and when it applies.

From OP-MC-PSS-RN Page 49 (Rev 43) 3.2 Abnormal and Emergency Operation 3.2.1 Abnormal Procedure APII or21A15500120 AP2O purpose, Cases, Symptoms, and basis for steps are covered in the AP Lesson Plan. I Objective#16 I 3.2.2 Blackout Alignment Blackout is a loss of power to the 4160 vac bus. When the low voltage condition is detected, the DIG will start and the sequencer will load the Blackout loads onto the bus. On receipt of a Blackout signal, Train A valves automatically assume low level alignment; Train B assumes SNSWP alignment. Many shared valves receive signals from both units to prevent loss of water from SNSWP. Isolation valves for all heat exchangers which are needed open automatically and the train related RN pump will start. All nonessential discharge is isolated except the containment vent units and NC pump motor cooler discharge. The containment vent units and the NC pump motor coolers are supplied with cooling water from A RN pump. The A RN pumps supply the containment ventilation units with cooling water because they have more NPSH since their suction is aligned to the LLI and because the RV pumps may not have power. Drawings 7.10 and 7.11 provides the unit blackout flow path. Drawings 7.12 and 7.13 provides the flow path for Train A and Train B Blackout respectively. If a Blackout occurs on the opposite unit, the non-blackout unit will have its non essential header isolated from the B RN pump as a result of RN41 B and RN43A closing (Refer to Drawing 7.5). In order to supply the non-essential header on the non blackout unit, the A Train RN pump must be started.

_

                                                                            -   0 UNIT#1 NC PUMP                                                                      PB COOLERS                                                                          I-UPPER CONT                                                        CD VENT UNIT ESSENTIAL               Cl) 0 NOR 1A RETURN                  C,,

CD B RN PUMP wç w -o ESSENTIAL C, fl NOR IA z o

w CD

                                                                            -o  CD CONTAINMENT w

NON-ESS NOR CD k 1%) . C) o ESSENTIAL 185 NOR lB 0 SB C) RV SYS RC CR DISCI1 OSSOVERl__j< 2BaAC a ESSENTIAL NOR 15 RETURN 297B

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 52 2552 APEO65 2.420 Loss of Instrument Air

PE065 GENERIC nowledge of the operational implications of EOP warnings, cautions, and notes. (CFR: 41.10/43.5/45.13) Given the following conditions on Unit 1:

A loss of offsite power has occurred
Both EDGs failed to start
ECA-0.0 (Loss of All AC Power) has been implemented
VI Header pressure is 20 PSIG Based on the conditions above, CA flow may have to be controlled locally to prevent Which ONE (1) of the following completes the statement above?

A. SG overfill B. CA pump runout C. lossofheatsink D. loss of shutdown margin Tuesday, July 13, 2010 Page 148 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 52 2552 eneral Discussion In ECA-0.0, VI header pressure is checked to be greater than 60 PSIG. If it is not, there is a CAUTION in the RNO for this step that states [çontrolling CA flow in the following enclosure is time critical to prevent SG overfill and loss of the TD CA pump. Answer A Discussion CORRECT: See explanation above. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because the loss of VI results in the CA control valves failing open. It is reasonable for the applicant to conclude that this would result inapump_runout condition. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant concludes that the CA flow control valves fail closed on a loss of VI. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because the uncontrolled CA flow will cause a cooldown and it is reasonable for the applicant to conclude that the cooldown could result in enough of a positive reactivity addition to cause a loss of shutdown margin. Basis for meeting the KA KA is matched because the applicant must be familiar with the CAUTION in the RNO for checking VI header pressure greater than 60 PSIG and why it isaconcern during ECA-0.0. Basis for Hi Cog Basis for SRO only Job Leve cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Learning Objective:

1) EPECAOO4

References:

1) ECA-0.0 APEO65 2.4.20 Loss of Instrument Air

APEO65 GENERIC Knowledge of the operational implications of EOP warnings, cautions, and notes. (CFR: 41.10/43.5/45.13) 401-9 Comments: RemarkslStatus 401-9 Comments: C is NP as written. PTS is way too a significant event for the initial conditions. Consider changing to C to CA pump run out Resolution / Comments: Developed revised question per Lead Examiners recommendation. Rearranged distracters for psychometrics. Distracter analysis will need work if revised question is acceptable. See attached file for revised copy of question. Tuesday, July 13, 2010 Page 149 of 294

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2010 MNS SRO NRC Examination QUESTION 52 Tuesday, July 13, 2010 Page 150 of 294

Question 52

References:

From ECA-O.O: MNS LOSS OF ALL AC POWER PAGE NO. EP/21A/5000/ECA-0.0 7 of 171 Rev. 31 UNIT 2 ACTIONISX?EOTED RETON5E RESPONSE NOT OBTAEO

10. Control intact SIG levels as follows:

Check N!R level in any intact S/G - a. Maintain maximum CA flow until N/R GREATER THAN 11% (32% ACC) level in at least one intact SIG is greeter than 11% (32% ACC) b Check VI header pressure - GREATER b. Perform the following: THAN 60 PSIG. CAUTION Controlling CA flow in the following enclosure is time critical to prevent SIG overfili and loss of TD CA pump.

1) IF CA flow cannot be throttled with CA control vaives in subsequent steps. THEN control flow PER EP/2A5000G-l (Generic Enclosures), Enclosure 16 (CA Flow Control With Loss of Vl)

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 53 APEO77 AK2.03 Generator Voltage and Electric Grid Disturbances

iowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: (CFR: 41.4, 41.5, 41.7, 41.10 I

.5.8)

Sensors, detectors, indicators Unit I & 2 are operating at 100% RTP:

The TCC has notified the Control Room that the Real Time Contingency Analysis (RTCA) indicates that switchyard voltage would NOT be adequate should a Unit Trip occur
The CR Supervisors implement API1IA/5500105 and API2IA/5500105, Generator Voltage and Electric Grid Disturbances
The OAC is not in service
Unit 1 Main Generator Voltage is 23.8 KV
Unit 2 Main Generator Voltage is 24.2 KV
Unit I & 2 Main Generator MWs are 1200
Unit 1 Main Generator MVARs are 450
Unit 2 Main Generator MVARs are 475
2 pressure on both generators is 75 PSIG H

Which ONE (1) of the following actions is required to be taken by the Unit 1 & 2 crews? REFERENCE PROVIDED A. Depress RAISE on the VOLTAGE ADJUST for Unit 1 ONLY. B. Depress RAISE on the VOLTAGE ADJUST for Unit 2 ONLY. C. Depress LOWER on the VOLTAGE ADJUST for Unit 1 ONLY. D. Depress LOWER on the VOLTAGE ADJUST for Unit 2 ONLY. Tuesday, July 13, 2010 Page 151 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 53 2553 C General Discussion For Unit 1, the 95 PERCENT (22.8kv) capability curve is used and it is determined that 450 MVAR is outside the limits of the curve which requires the VOLTAGE ADJUST to be lowered to reduce the lagging MVARs. For Unit 2, the 100 PERCENT (24kV) capability curve is used and it is determined that 475 MVAR is within the limits of the curve. Therefore, MVARs do NOT need to be adjusted on Unit 2. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant chooses the wrong capability curve to read or misreads the correct curve and does not understand how the operation of the VOLTAGE ADJUST affects unit MVARS. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant chooses the wrong capability curve to read or misreads the correct curve and does not understand how the operation of the VOLTAGE ADJUST affects unit MVARS. - Answer C Discussion CORRECT: See explanation above. Answer D Discussion INCORRECT: See explanation above. -_________________________________________________ PLAUSIBLE: This answer is plausible if the applicant chooses the wrong capability curve to read or misreads the correct curve. Basis for meeting the KA KA is matched because a Grid Disturbance has occurred and the applicant must use the indications provided to determine the impact on Main Generator operation. ( isis for Hi Cog his is a higher cognitive level question. First the applicant must use the indications provided to determine which capability curve to use. The

applicant must then determine if each of the generators is operating within the limits of their respective capability curves. The applicant must then chose the appropriate action based on whether the unit is operating within the limits of the curve. Since this question requires multiple mental steps, it is a higher cognitive level question. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK MNS Exam Bank Question #AP05012 Development References Student References Provided Learning Objectives: Unit 1 & 2 Generator Capability Curves

1) N/A

References:

1) AP/1/A15500/05 and AP/2/A!5500/05 Generator Voltage and Electric Grid Disturbances APEO77 AK2.03 Generator Voltage and Electric Grid Disturbances

Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: (CFR: 41.4, 41.5, 41.7, 41.10/ 45.8) Sensors, detectors, indicators 1-9 Comments: Remarks/Status 401-9 Comments: The stem requires an action. Distractor A is NOT an action and therefore is NP. Consider using raise for A and B in some Tuesday, July 13, 2010 Page 152 of 294

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2010 MNS SRO NRC Examination QUESTION 53 2553 fashion. Resolution / Comments: Revised answers A and B to include raise. Added if any to stem. See attached copy of revised question. Tuesday, July 13, 2010 Page 153 of 294

Question 53 Parent Question: From AP111A15500105: GENERATOR VOLTAGE AND ELECTRIC GRID MNS PAGE NO DISTURBANCES AP/1/N5500105 5 of 30 Rev. 007 UNIT 1 ACTION/EXPECTEP RESPONSE RESPONSE NOT OBTAINEP

7. Monitor Generator Capability Curve as follows:

NOTE In the following step, if Generator voltage is fluctuating above and below 24 Ky, then assume voltage is less than 24 Ky.

a. Check Generator voltage LESS THAN

                                           -                       a. Perform the following:

24 Ky.

1) Monitor Generator Capability Curve I PER Enclosure 1 (Generator Capability Curve 24 Ky).

_2) GOTOStep8.

b. Check OAC - IN SERVICE. b. Perform the following:

1) Monitor Generator Capability Curve PER Enclosure 2 (Generator Capability Curve 22.8 Ky).

2) GOTOStep8.

c. Monitor Generator Capability Curve PER OAC turn on code GENCAP.

8. Check Generator MVARs WITHIN - GO TO Step 11.

LIMITS OF GENERATOR CAPABILITY CURVE.

9. IF AT ANYTIME capability curve exceeded, THEN perform Steps 11 and 12.

l0. GOTOStepl3.

GENERATOR VOLTAGE AND ELECTRIC GRID MNS PAGE NO AP/1IA/5500/05 DISTURBANCES 6 of 30 Rev. 007 UNIT 1 A::T:O:*:/E:p:ITED RESPONSE RESPONSE NOT OBTAINER

11. Adjust MVARs to within the capability curve by performing one of the following:

Depress LOWER on the VOLTAGE ADJUST to reduce lagging MVARs OR

  . Depress RAISE on the VOLTAGE ADJUST to reduce leading MVARs.

12. Check Generator MVARs WITHIN

                                -                  ) actions in Step 11 do not restore LIMITS OF GENERATOR CAPABILITY                 MVARs, THEN perform the following:

CURVE.

a. IF voltage regulator in AUTO, THEN perform the following:

1) Place voltage regulator in MAN.

2) Adjust MVARs to within the capability curve.

b. IF unable to maintain MVARs within curve, THEN remove generator from service as follows:

1) IF greater than P-8, TH.N perform the following:

a) Trip reactor. b) GOTOEPJ1/A15000/E-0 (Reactor Trip or Safety Injection).

2) IF less than P-8, THf{ perform the following:

a) Trip turbine. b) QTQAP/1/A/55Q0/02 (Turbine Generator Trip).

From AP121A15500105: MNS GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES PAGE NO API2JAI5500/O5 4 of 25 UNIT 2 Rev. 4 I I I ::z:zz:,:::::::::::

7. Monitor Generator Capability Curve as follows:

NOTE In the following step. if Generator voltage is fluctuating above and below 24 KV, then assume voltage is less than 24 Ky.

a. Check Generator voltage LESS THAN

                                          -                         a. Perform the following:

24 Ky.

1) Monitor Generator Capability Curve I PER Enclosure 1 (Generator Capability Curve 24 Ky).

2) GOTOStep8.

b. Check OAC - IN SERVICE. b. Perform the following:

1) Monitor Generator Capability Curve PER Enclosure 2 (Generator Capability Curve 22.8 Ky).

2) GOTOStep8.

c. Monitor Generator Capability Curve PER OAC turn on code GEN CAP.

8. Check Generator MVARs WITHIN - G T Step 11.

LIMITS OF GENERATOR CAPABILITY CURVE. 9, If AT ANYTIME capability curve exceeded, iMN perform Steps 11 and 12.

10. cTStep13.

D MVARS

             &               - N) C.)             Q -

0 0 0 0 0 0 C C C C C C 0 C C C C C C C C C C C C C C C C

-<

ci 0 C) rn>O ZOC m

                                                              -i C)

Z m m

                                                         -I 0

D MVARS

            -  )   .. cJ - o <

a a z cD c cD c, c c c c c c, Q a

r m

                                          ;    0 0

m rn om omc>F OUG)m O0tIm..< I-

D>mO m a-I

MVARS 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 C 0 zVcm momQO

C m < m 0

MVARS

C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C C Q 0

                                                      - 0 0cnm..<

rz

OW 1IA/6100!022 Enclosure 4.3 UNIT I Table 3.1.4 MeGuire Nuclear Statiozi Generator CapabilIty Curve Application Guidance 22.8kv Capability Curve (Curve 3.1.1 of Enclosure 4.3) to be used when generator output 22.8kv and 24.0 kV. voltage is between 2 24.0kV Capability Curve (Curve 3.1.2 of Enclos ure 4.3) to be used when generator output voltag higher. e is 24.0kv or NOTE: MVAR limits provided in Enclosure 4.3 are based upon Pull Power (1200 MWs) operati power MVAR limits should be obtained from genera on. At reduced tor capability curves. Actual MVAR limits operating generator voltage, 142 pressure, MW output are based upon

                                                        , etc.

3 When generator MVARs exceed capability curve, refer to AP / I IA! 5500! 005 (Generator Voltage and Electric Grid Disturbances) UNIT 1

Question 53 Parent Question: AP050121 pt Unit 1 & 2 are operating at 100% RTP:

The TCC has notified the Control Room that the Real Time Contingency Analysis (RTCA) indicates that switchyard voltage would not be adequate should a Unit Trip occur.
The CR Supervisors implement AP111A155001005 and AP/2/A15500/005, Generator Voltage and Electric Grid Disturbances.
Enclosure 1 (Abnormal Generator or Grid Voltage) is implemented.
Step 1 of Enclosure 1 directs the operators to Check Generator TIED TO

GRID

Unit 1 Main Generator Voltage is 24.2 KV
Unit 2 Main Generator Voltage is 23.9 KV
Unit 1 & 2 Main Generator MWs are 1200
Unit I Main Generator MVARs are 475
Unit 2 Main Generator MVARs are 450 Which one of the following actions are required to be taken by the Unit 1 & 2 crews?

A. Monitor Unit 1 & 2 MVARs and continue with the procedure B. Place the voltage Regulator in Manual and adjust MVARs on both units C. Place the voltage Regulator in Manual and adjust MVARs on Unit 1 only D. Place the voltage Regulator in Manual and adjust MVARs on Unit 2 only Answer 576 D

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 54 WEO4 EK1.3 LOCA Outside Containment

riowledge of the operational implications of the following concepts as they apply to the (LOCA Outside Containment) FR: 41.8 / 41.10, 45.3) Annunciators and conditions indicating signals, and remedial actions associated with the (LOCA Outside Containment). Given the following conditions on Unit 1:

A Reactor Trip and SI have occurred due to low Pressurizer pressure
Crew is performing the actions of E-0 (Reactor Trip or SI)
SI termination criteria cannot be met at this time
Containment parameters are normal
Both ND pumps are tripped
FWST level indicates 340 inches
1EMF-1 (ND Area Monitor) is in Trip 2 at 1.5E2 mREM/hr
1EMF-41 (Aux Bldg Ventilation) is in Trip 2 alarm Based on the above indications, the crew will transition to (1) and the strategy implemented to mitigate this event is (2)

Which ONE (1) of the following completes the statement above? A. 1. ECA-1 .1 (Loss of Emergency Coolant Recirculation)

2. to identify and isolate the break B. 1. ECA-1 .1 (Loss of Emergency Coolant Recirculation)

2. to delay depletion of the FWST by reducing oufflow and initiating makeup C. 1. ECA-1 .2 (LOCA Outside Containment)

2. to identify and isolate the break D. 1. ECA-1 .2 (LOCA Outside Containment)

2. to delay depletion of the FWST by reducing outflow and initiating makeup Tuesday, July 13, 2010 Page 154 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 54 General Discussion - In the scenario described in the stem of this question, the applicant is presented with a set of indications which would require a transition directly from E-0 to ECS 1.2. This transition would be based on the fact that SI termination cannot be met, FWST inventory is being depleted and there are multiple indications of elevated radiation levels in the Aux building with containment conditions normal. Step 39 of E-0 would direct this transition. trategy employed by ECA l.2(LOCA Outside Containment) is to identify and isolate the leak. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because the applicant is presented with conditions which meet the entry criteria for ECA -1.1 (Loss of ECR). FWST inventory is being depleted without a corresponding increase in containment sump level. This would be a correct answer if a transition were not being made directly from E-0. Part 2 plausible because ECA 1.1 contains actions to reduce the loss of inventory and initiate make up to restore FWST level. It is plausible the applicant would misinterpret the actions to_identify_and isolate the break as one of those action because it would reduce the loss of inventory. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because the applicant is presented with conditions which meet the entry criteria for ECA -1.1 (Loss of ECR). FWST inventory is being depleted without a corresponding increase in containment sump level. This would be a correct answer if a transition were not being made directly from E-0. Part 2 is correct strategy for the procedure given in this distracter and therefore plausible. - - Answer C Discussion CORRECT: See explanation above. nswer D Discussion CORRECT: See explanation above. PLAUSIBLE: Part 1 is correct and therefore plausible. Part 2 is plausible because actions contained in ECA 1.2, if successful would delay depletion of the FWST by reducing outflow. Basis for meeting the KA KA is matched because the applicant must evaluate annunciators and indications associated with implementation ECA-I.2 (LOCA Outside Containment). The operational implication of the given indications would be the transition to the procedure. He must then identify the remedial actions associated with performing ECA-1. 2. 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, solve a problem by selecting the correct procedure and identifying the correct strategy associated with that transition. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Learing Objective:

1) EP-El #4 E-l Lssson Plan A 1.2 (LOCA Outside Containment) - - -

WEO4 EK1 .3 LOCA Outside Containment

Knowledge of the operational implications of the following concepts as they apply to the (LOCA Outside Containment) (CFR: 41.8 /41.10, 45.3) Annunciators and conditions indicating signals, and remedial actions associated with the (LOCA Outside Containment). Tuesday, July 13, 2010 Page 155 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 54 Z9rnents: Remarks!Status 40 1-9 Comments: The stem requires an action to mitigate the event. Delaying depletion of the FWST is not an action to mitigate a LOCA

outside containment event. Isolating the break is ALWAYS a good thing. Distractors B2 and D2 are NP. Replace these 2 distractors. This Q is U because of 2 NP distractors. Resolution I Comments: Added Both ND pumps are tripped to stem of question to give plausibility to A and B. Replaced should with must in the stem of the question per Lead Examiners General Comments. Tuesday, July 13, 2010 Page 156 of 294

Question 54

References:

From Lesson E.O: MNS RE.CTCR TRIP DR SAFETY INJECTIOk PAGE NO. ER?1Lj5J3DiE-O 28 of 27 Re ° 1NIT1 1 LLLLLW.LL I .14 I 4! LL LI.. LLL L

39. Check for potential leak in aux bldg:

a. Che;k aux bdQ radiatrn: a. Peorm the foltown2:

_: Al area nicnctorEMFs NCRr.AL

                                          -                          1) Detemiine location of actiity usirt; any of the follot.lng:

I ElF-4i (A;ix 611g erti ton

FdF nhrnnn OAC :trn nn cole E?1Fi _: EMF4 sample point realm; the hi;liest on OAC him or oole ESF-Ai) _: Area mDnitor EMF alorms.

2) Dispatcn operator tD locate ant icoiae pc>:entiol Lnit 1 3: IF cause of alami is LOCA outside ontanniert. 1HEI\ GO TO EP:iA5CDDECA-L2 i:LDC:.S

                                                                         )utsdc Contc imcnt;.

NOTE T efoilowtnci step is checking for a si;lnificant NC leak into the ND Ss:em.

h. Che:k NC to ND pessure hu.ar S. W ND Ssten1 paraners indcate intact as follows: LOCA outside containment. THEN GO TO EP/1.A/5EOO/ECA-l.2 (LOC.A ND Temperature - NORMAL Outide Containment;.

ND Ffow NORMAL

I ND Pressure - MDRh1AL

From E-1 Lesson Plan (Page 191): 1.0 ECA-1.1, LOSS OF EMERGENCY COOLANT RECIRCULATION 1.1. Purpose This procedure provides instruction for when emergency coolant recirculation (ECR) capability is lost. Loss of ECR is defined as the loss of the ability to provide the recirculation function following a LOCA, i.e., the loss of the ability to inject fluid from the sump to the NC using a ND pump. The objective of the loss of ECR procedure is threefold:

1) To continue attempts to restore emergency coolant recirculation capability,

2) To delay depletion of the FWST by adding makeup fluid and reducing oufflow, and

3) To depressurize the NC to minimize break flow and cause S/I accumulator injection.

From E-1 Lesson Plan (Page 191): 2.0 ECA-1.2, LOCA OUTSIDE CONTAINMENT

21. Purpose This procedure provides guidance for a LOCA that occurs outside containment.

Specifically, the objective of this procedure is to provide actions to identify and isolate a LOCA outside containment. This entire EP is a significant deviation from the ERGs. Isolating an ISLOCA into the ND system is considered PRA significant operator action as described in PIP M02-247. The valves used to do this isolation (Nl-173A1178B) are not designed to close against the DP that could be seen during an ISLOCA, since this is a beyond design basis event. To meet the intent of this EP to isolate a break on low pressure ND system piping, this EP includes actions to cooldown and depressurize the NC system to the point where the isolation valves are capable of closing. 2.2. Symptoms!Entry Conditions ECA-1 .2 is entered when either of the following conditions occur:

1. In E-O, when abnormal radiation occurs in the Aux Building due to a loss of NC system inventory outside containment.

2. When it is determined in E-1 that the cause of abnormal radiation is due to a loss of NC inventory outside containment.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 55 WEO5 EK3.2 Loss of Secondary Heat Sink

nowledge of the reasons for the following responses as they apply to the (Loss of Secondary Heat Sink) ,JR: 41.5/41.10, 45.6, 45.13) Normal, abnormal and emergency operating procedures associated with (Loss of Secondary Heat Sink). Given the following conditions on Unit 1:

A medium-break LOCA occurred in Containment
Containment pressure peaked at 2.7 PSIG and is slowly decreasing
The crew has implemented FR-H.1 (Response to Loss of Secondary Heat Sink)
All attempts to restore flow to the S/Gs from the CA system have been unsuccessful

1. Based on these conditions, the NC pumps must be stopped to

2. The EARLIEST time (based on S/G conditions) that the crew is required to establish NC system bleed and feed is when W/R level in at least 3 S/Gs is less than Which ONE (1) of the following completes the statements above?

A. 1. prevent NC pump impeller damage due to low pressure operation

2. 24%

B. 1. prevent NC pump impeller damage due to low pressure operation

2. 36%

C. 1. conserve secondary inventory by reducing NC system heat input

2. 24%

D. 1. conserve secondary inventory by reducing NC system heat input

2. 36%

Tuesday, July 13, 2010 Page 157 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 55 2555 General Discussion Per FR-H. I Basis Document the basis for Step 9 states: STEP 9 Stop all NC pumps. PURPOSE: To stop NC pumps in order to extend the time to restore feed flow to the S/Os. BASIS: NC pump operation results in heat addition to the water in the NC system. By tripping the NC pumps, the effectiveness of the remaining water inventory in the S/Os is extended, which extends the time at which the operator action to initiate feed and bleed must occur. This extension of time is additional time for the operator to restore feedwater flow to the S/Os. Per FR-Hi, if WR Level in at least 3 SGs is less than 24% (36% ACC), Feed and Bleed of the NC system is initiated. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because it is reasonable for the applicant to conclude that during a LOCA the reduction in NC system subcooling and potential voiding in the NC system could result in a scenario where NC pump cavitation could occur. Part 2 is correct. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because it is reasonable for the applicant to conclude that during a LOCA the reduction in NC system subcooling and potential voiding in the NC system could result in a scenario where NC pump cavitation could occur. rt 2 is plausible because this is this is the adverse containment condition value for initiating feed and bleed. However, adverse condition quirements_have not been met. Answer C Discussion CORRECT: See explanation above. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: Part I is correct. Part 2 is plausible because this is this is the adverse containment condition value for initiating feed and bleed. However, adverse condition requirements have not been met. Basis for meeting the KA -________________________________________________________________________ KA is matched because the applicant must demonstrate a knowledge of the emergency operating procedure associated with Loss of Secondary at Sink (FR-H. 1) with regards to the mitigative strategy in the procedure and the basis for performing major actions in the procedure. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK Neve AUDIT QO (Bank 1028) LDevelopment References Student References Provided srning Objective:

EP-FRH Objective 4

References:

1) EP-FR-H.1 Loss of Secondary Heat Sink

2) FR-H. I Background Document Tuesday, July 13, 2010 Page 158 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 55 WOG Background_Document for FR-H.l

 £05 EK3.2 Loss of Secondary Heat Sink
              -

Knowledge of the reasons for the following responses as they apply to the (Loss of Secondary Heat Sink) (CFR: 41.5 /41.10, 45.6, 45.13) Normal, abnormal and emergency operating procedures associated with (Loss of Secondary Heat Sink). 401-9 Comments: RemarkslStatus 40 1-9 Comments: Al and RI are borderline NP because seal leak off is so insignificant with respect to a LOCA unless the seal is blown. Consider changing both to prevent NC pump impeller damage to low pressure operations or something equivalent. This Q is U because of 2 NP distractors. Resolution I Comments: Revised per Lead Examiners recommendation. Revised distracter analysis to match new answers A and B. See attached file for proposed revision. - Tuesday, July 13, 2010 Page 159 of 294

Question 55

References:

From Lesson Plan OP-MC-EP-FRH: STEP 8 Check steam dumps: PURPOSE: Place steam dumps in pressure mode of control before stopping all NC pumps. BASIS: Provide better control of steam dumps under natural circ conditions. If no NC pump is running, then the NC system average temperature will be higher than the no-load value as natural circulation conditions are established. However, if the steam dump system is working properly, the cold leg temperatures will stabilize at the no-load value. STEP 9 Stop all NC pumps. PURPOSE: To stop NC pumps in order to extend the time to restore feed flow to the SIGs. BASIS: NC pump operation results in heat addition to the water in the NC system. By tripping the NC pumps, the effectiveness of the remaining water inventory in the SIGs is extended, which extends the time at which the operator action to initiate feed and bleed must occur. This extension of time is additional time for the operator to restore feedwater flow to the SIGs.

From WOG Background Document for FR-H.1: 2.5 Reactor Coolant Pump Operation Operation of reactor coolant pumps will affect the dryout time of the steam generators due to RCP heat addition and, therefore, will affect the time at which operator action to initiate bleed and feed must occur. Studies have been performed using the LOFTRAN code (Reference 2) to assess the impact of RCP operation on the time PORVs will open without operator action and the time to steam generator dryout for a loss of main feedwater event without AFW available. A four-loop plant typical of current Westinghouse design was used. It had a core power of 3411 Mwt and an RCP steady state power of 14 Mwt. Model F steam generators were also assumed. Thus, while this plant is not identical to the one used in References 1, 3 and 4, the study will be representative of Westinghouse plant response and sufficient to determine the impact of RCP status on the time available before operator action to initiate bleed and feed is required. The cases analyzed were: Case 1: RCPs running throughout transient Case 2: RCPs tripped at reactor trip Case 3: RCP5 tripped 5 minutes after reactor trip The focus of the analysis was to determine the additional time available to the operator as a result of eliminating RCP heat from the system before action to initiate bleed and feed became necessary. Thus, the time of two events was used to determine the impact of RCP trip time. The two events are 1) the time when PORVs automatically open as a result of the degraded heat transfer capability of the steam generator and 2) the time when steam generator secondaries dry out. Table 1 shows a comparison of the three cases. Case 1 represents a situation where steam generators would experience the earliest dryout due to the RCP heat load and Case 2 is where the steam generators would experience the latest dryout. The extension in dryout time from Case 1 to Case 2 is between 7

and 9 minutes, depending upon the indication of dryout that is chosen. The use of the time to PORV opening will have some uncertainty due to the uncertainty in predicting non-equilibrium effects in the pressurizer. However, PORV opening time is probably the best indicator obtainable from the LOFTRAN analysis of the time available until bleed and feed must be initiated. Case 3, where the RCPs are tripped 5 minutes after reactor trip, is a best estimate expectation of when the operator can be expected to trip RCP5 following a reactor trip based on guidance provided in this guideline. Thus, the extension in time to loss of heat sink symptoms is the most realistic that could be expected based on anticipated operator response. The extension to loss of secondary heat sink symptoms is about 5 minutes based on PORV opening time. This compares favorably with the extension already seen between Cases 1 and 2. Thus, operator action to trip RCPs upon entering this guideline for loss of secondary heat sink can appreciably delay the need for bleed and feed and the loss of secondary heat sink. Thus, time can be gained for the operator to establish a means of supplying feedwater. Delaying the loss of secondary heat sink is not the only reason for tripping RCPs. RCPs running can also reduce the effectiveness of bleed and feed. RCP heat input to the RCS will result in increased steam generation hindering the depressurization of the RCS during bleed and feed. The higher pressure produced by RCP operation will reduce SI flow and increase inventory lost through the PORVs. Therefore, RCPs should be tripped if AFW flow cannot be established immediately after entering this guideline.

From FR-H.1: MNS RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE NO EP! I !NS000IFR-H. 1 3 of 98 UNIT 1 Rev. 14 ADMON/EXPECTE: P2SIDNEE REEONSE NOT c3TA:KED

4. Chect at least one of the following NV Q IQ Step 20.

pumps AVAILABLE:

IA NV pump OR
ID NV pump.

5. Chec if NC System feed £Iid bleed should be initiated:

a. Check W/R level in at least 3 S!Gs - a. Perform the following:

LESS THAN

1) Monitor feed and bleed initiation criteria.

2) WHEN criteria satisfied, THEN GO JQ Step 20.

_3) .QjQ StepS. _b. 3QIQStep20.

6. Ensure S/G BB and NM valves CLOSED PEREnclosure3(SfGBBand Sampling Valve Checklist).

7. Attempt to establish CA flow to at least one S/C as follows:

a. Check puwei Lu both MDCA puiii - a. Peilunit Lite Iollowiiiy.

AVAILABLE.

                                                         . IE lETA lETS deenergized, THEN restore power to the affected essential bus PER APIIIAJSSOO1C7 (Loss of Electftal Power).
                                                         . IF the essential bus is energized, THEN dispatd operator to determine cause of breaker failure.

b. Ensure control room CA valves aligned PER Enclosure 4 (CA VJve Alignment).

c. Sat all available CA pumps.

MNS RESPONSE TO LOSS OF SECONDARY HEAT SINK PAGE ND. EP!! ?ASOOOIFR-H. I 25 of 98 UNIT 1 Rev. 14 ACTIDN/EXPZDTED a:sr ONSE EEPCNSE NOT OETANEI

19. Check if NC System feed and bleed should be initiated:

a. Check feed and bleed HAS BEEN

                                  -                a. Q JQ Step 1 9.c.

PREVIOUSLY ESTABLISHED PER STEPS 21 through 25. _b. .QjQStep36.

c. Check WR level in at least 3 S?Gs - c. RETURN JQ Step 1.

LESS THAN 24% (36% ACC).

20. Perform Steps 21 tIirouqh 25 quickly to establish NC heat removal by NC feed and bleed..

21. Ensure all NC pumps OFF.

22. Initiate S/I.

Parent Question (Bank Question 1028): Unit 2 was operating at 100% power. Given the following:

A medium break LOCA occurred in containment
Containment pressure peaked at 2.7 psig and is slowly decreasing
The crew has implemented EP/2/A/5000/FRH.1 (Response to Loss of Secondary Heat Sink)
All attempts to restore flow to the S/Gs from the CA system have been unsuccessful

1. Which one of the following identified the next source of feed water that EP/2/A/5000IFR-H. 1 will prioritize for restoration of flow to the S/Gs?

2. What is the earliest time (based on S/G conditions) the crew is required to establish bleed and feed?

A. 1. Through the CM/CF system using a main feed water pump

2. When W/R level in at least 3 S/Gs is less than 24%

B. 1. Through the CM/CF system using a main feed water pump

2. When W/R level in at least 3 S/Gs is less than 36%

C. 1. Through the CM/CF system using a hotwell and booster pumps only

2. When W/R level in at least 3 S/Gs is less than 24%

D. 1. Through the CM/CF system using a hotwell and booster pumps only

2. When W/R level in at least 3 S/Gs is less than 36%

ANSWER: A

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 56 L6 EAI.1 - Loss of Emergency Coolant Recirculation bility to operate arid for monitor the following as they apply to the (Loss of Emergency Coolant Recirculation) JR: 41.7/45.5/45.6) Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features. Given the following conditions on Unit 1:

EPI1IAI5000IECA-1.1, Loss of Emergency Coolant Recirculation, has just been implemented
The FWST LEVEL LO-LO alarm is LIT Which ONE (1) of the following actions are required FIRST?

A. When FWST level decreases to less than 20 inches, reset Containment Spray AND stop the NS pumps. B. When FWST level decreases to less than 20 inches, stop the NS pumps ONLY. C. Immediately reset Containment Spray AND stop the NS pumps. D. Immediately stop the NS pumps ONLY. Tuesday, July 13, 2010 Page 160 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 56 C General Discussion In accordance with ECA-l.l Loss of Emergency Coolant Recirculation Enclosure I (Foldout) if FWST level goes below the FWST LEVEL LO LO alarm setpoint (33 inches) and the NS pumps are taking a suction from the FWST then Reset Containment Spray and Stop both NS pumps. Additionally, if FWST level goes below 20 inches then stop ALL pumps taking a suction on the FWST. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because the correct action is to reset Containment Spray and stop the NS pumps. However, it should be done immediately since FWST level is below the FWST LEVEL LO-LO alarm setpoint. The FWST level of 20 inches is plausible because at that level ECA-1.l directs all pumps taking suction from the FWST to be stopped. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant does not recall that Enclosure 1 directs resetting Containment Spray prior to stopping the NS pumps. The FWST level of 20 inches is plausible because at that level ECA-1.l directs all pumps taking suction from the FWST to be stopped. Answer C Discussion CORRECT: See explanation above. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant does not recall that Enclosure 1 directs resetting Containment Spray prior to stopping the NS pumps. The NS pumps should be stopped immediately since FWST level is below the FWST LEVEL LO-LO alarm setpoint. However, it ould be done AFTER Containment Spray is reset. asis for meeting the KA The KA is matched because the applicant demonstrates the ability to operate the NS pumps (i.e. Ability to operate and I or monitor components as they apply to the Loss of Emergency Coolant Recirculation) with regards to knowing what conditions in ECA-l.l require the NS pumps to be stopped. Basis for Hi Cog sis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK 2005 NRC Q17 (Bank 421) Development References Student References Provided Learning Objectives:

1) EPE1005

References:

1) EP/l/A15000/ECA-1.l WE1 1 EAI .1 Loss of Emergency Coolant Recirculation

Ability to operate and I or monitor the following as they apply to the (Loss of Emergency Coolant Recirculation) (CFR: 41.7145.5/45.6) Thmponents, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and riual features. 401-9 Comments: RemarkslStatus 401-9 Comments: Tuesday, July 13, 2010 Page 161 of 294

___________ FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION f 56 Since A and B are both actions in ECA 1.1, the WOOTF statement should include the word FIRST to completely rule out A and B. Facility please reevaluate. Resolution / Comments: Reworded stem of question to include FIRST per Lead Examiners recommendation. See attached file for revised copy of question. Tuesday, July 13, 2010 Page 162 of 294

Question 56

References:

From EPII/AI5000IECA-1 .1: MNS LOSS OF EMERGENCY COOLANT RECIRC PAGE NO. EP/1/A15000/ECA-1 .1 of 1Q4 Enclosure 1 Page 1 of 1

UNIT 1 Foldout R ev.

1. Emergency Coolant Recirc Capability Restoration:

WHEN Cold Leg Recirc capability is restored, THEN GO TQ.. Step 6.f in body of this procedure.

2. ECCS Suction Monitoring Criteria:

LF FWST level goes below FWST LEVEL LO-LO alarm setpoint (33 inches). ANNS pumps are taking suction from the FWST. THE:

a. Reset Containment Spray.

b. Stop both NS pumps.

if FWST level goes below 20 inches. THEN stop all pumps taking suction from the FWST.
IF suction source is lost to any NV, NI, ND, or NS pump. THEN stop pump.

3. CA Suction Sources:

IF CA Storage Tank (water tower) goes below 1.5 ft, THEN perform EP/1/A15000/G-1 (Generic Enclosures), Enclosure 20 (CA Suction Source Realignment).

4. CLA IsolatIon:

IF at least two NC T-Hots are less than 354F, THEN isolate CLAs PER Enclosure 10 (CLA Isolation).

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2005 CNS SRO NRC Examination QUESTION 17 421 WE! 1 EA2.2 Loss of Emergency Coolant Recirculation

Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation) (CFR: 43.5 /45.13) Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments. Given the following:

EPIIIAI5000IECA-1 .1, Loss of Emergency Coolant Recirculation, has just been implemented
Refueling Water Storage Tank (FWST) level is 4.5%

Which of the following procedure actions is performed flt while attempting to restore recircu lation? A. Initiate makeup to the FWST. B. Start one reactor coolant pump. C. Makeup to the NC system from the standby makeup pump. D. Secure all ECCS and NS pumps taking a suction from the FWST. Tuesday, July 13, 2010 Page 33 of 202

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2005 CNS SRO NRC Examination QUESTION 17 42l General Discussion All these actions are done, but are not done first. Based on stem conditions, Enclosure 1 applies which required all pumps taking a suction from the FWST to be secured. This is also a step in the body of the procedure. Answer A Discussion Answer B Discussion Answer C Discussion Answer D Discussion Basis for meeting the KA Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK Development References Student References Provided LessonE OP-CN-EP-EP2 ObjectivesEi2O REFERENCES EEP/1/A15000/ECA-l .1 WEll EA2.2 Loss of Emergency Coolant Recirculation

Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation) (CFR: 43.5/45.13) Adherence to appropriate procedures and operation within the limitations in the facility*s license and amendments. 401-9 Comments: RemarkslStatus Tuesday, July 13, 2010 Page 34 of 202

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 57 2557 AK2.04 Emergency Boration

( Knowledge of the interrelations between Emergency Boration and the following: (CFR 41.7 /45.7) Pumps Given the following conditions on Unit 1:

An ATWS has occurred
The crew has entered FR-S.1 (Response to Nuclear Generation/ATWS)
During the initiation of emergency boration, the following indications are noted:

o Charging Flow = 47 GPM o Letdown Flow=75GPM o NC system pressure is 2300 PSIG o 1A NV pump is ON with suction aligned to the VCT o IA and 1 B BAT pumps are ON o 1NV-265B (Boric Acid To NV Pumps) is open o INV-244A (Chrg Line Cont Isol) is open o 1NV-245B (Chrg Line Cont Isol) is open In accordance with FR-Si, the MINIMUM required emergency boration flow is (1) and if that flow is NOT met the Operator will (2) Which ONE (1) of the following completes the statement above? A. 1. 3OGPM

2. increase charging flow B. I. 6OGPM

2. increase charging flow C. 1. 3OGPM

2. align the NV pump suction to the FWST D. I. 6OGPM

2. align the NV pump suction to the FWST Tuesday, July 13, 2010 Page 163 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 57

.jeneral Discussion In accordance with FR-S. 1 there must be a minimum of 30 GPM emergency boration flow to ensure adequate boric acid is getting to the reactor.

As part of the checks to ensure an adequate flow path an NV pump is started, both BA pumps are started, charging flow is checked to be greater than BA flow, NC system pressure is checked less than 2335 PSIG, and a check for adequate flow path is made. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is correct. Part 2 is plausible if the applicant concludes that increasing charging flow will also in turn allow emergency boration flow to increase (by reducing the back-pressure on the emergency boration line). Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because it is the normal BA flow that would be seen during a boration. Part 2 is plausible if the applicant believes that increasing charging flow will also in turn allow emergency boration flow to increase (by reducing the back-pressure on the emergency boration line). Answer C Discussion CORRECT: See explanation above. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because it is the normal BA flow that would be seen during a boration. Part 2 is correct. asis for meeting the KA he KA is matched because the applicant must know the interrelation between the Charging Pumps and Boric Acid pumps during an emergency boration and the action to be taken if the required flow from the boric acid pumps is not achieved. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Learning Objectives:

1) EPFRSOO3

References:

LS.1___ APEO24 AK2.04 Emergency Boration

Knowledge of the interrelations between Emergency Boration and the following: (CFR 41.7/45.7) Pumps Comments: RemarkslStatus 401-9 Comments: Stems should not include a should. I did NOT find it anywhere in the procedure. The word should be shall or will or must Tuesday, July 13, 2010 Page 164 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 57 2557 Resolution / Comments: Changed should to will per Lead Examiners recommendation. See attached file for revised question. Tuesday, July 13, 2010 Page 165 of 294

Question 57

References:

From FR-S.1: MNS RESPONSE TO NUCLEAR POWER GENERATIONJATWS PAGE NO. EP,2/A,50OOFRS. 1 3 of 29 Rev. ID UNIT 2 DTIcNzxTE:TTr R9CN RE5DNS NOT OBOAIND

5. Initiate emergency boration of NC System as follows:

a. Ensure one NV pump ON. - a. Place PD pump in service PER EPJ2IA/5000/G-l (Generic Enclosures), Enclosure 17 (PD Pump startup).

b. Align boretion flowpath us follows:

1) Open 2NV-265B (Boric Acid To NV Pumps).

2) Start both boric acid transfer pumps.

3) Check emergency boration flow - 3) IF NV pump suction is aligned to GREATER THAN 30 GPM. VCT, THEN align to FWST as follows:

a) Open 2NV-22IA (NV Pumps Suct From FWST). b) Open 2NV-222B (NV Pumps Suet From FWST). C) Close 2NV-l4lA (VCT Outlet Isol). d) Close 2NV-142B (VCT Outlet sol).

A.1TION/EXECTZD RESPCNSZ D,ZSPOJ5E NOT OWrflNED

5. (Continued)

a. Check if NV flowpath aligned to NC a. Perform the following:

System:

1) IF NV pump suction is nligned to

      . 2NV-244A (Charging Line Cant             VCT, ]EHKN align to FWST as Outside Isol) OPEN
                      -                           follows:

2NV-2456 (Charging tine Cant a) Open 2NV-221A (NV Pumps Outside Isol) OPENS

Suct From FWST). b) Open 2NV-222B (NV Pumps Suct From FWST). a) Close 2NV-141A (VCT Outlet Isol). d) Close 2NV-14213 (VCT Outlet Isol).

2) Open the following valves:

                                              . 2Nl-9A (NC Cold Leg nj From NV)

                                                  . 2Nl-iOD(NC Cold Leg In) From NV).

_3) GOTOStep 5.e.

d. Ensure charging flow is greater than emergency boration flow.

e. Check Pzr pressure LESS THAN

                          -                e. Perform the following:

2335 P81(3.

1) iF all Pzr PORVs and isolation valves open, THEN Q]Q Step ft

2) IF Pzr PORV(s) OR isolation valves closed, THEN open Pzr PORV(s) and isolation valves as required to reduce Pzr pressure less than 2135 PSIG (200 PSIG less than PORV auto open setpoint).

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 58 E8 APEO28 AK1.Ol Pressurizer (PZR) Level Control Malfunction

Knowledge of the operational implications of the following concepts as they apply to Pressurizer Level Control Malfunctions: (CFR 41.8/ l.lOI45.3) PZR reference leak abnormalities Given the following conditions on Unit 1:

Pressurizer Level transmitter 1 has failed low
Prior to removing the Level Channel 1 from service, a leak develops on the reference leg for Pressurizer Level transmitter 2 Based on these conditions, the indication for Pressurizer Level Channel 2 fails (1)

AND the Pressurizer Level Master Controller (2) Which ONE of the following completes the statement above? A. 1. low

2. swapsto MANUAL B. 1. low

2. remains in AUTO C. 1. high

2. swaps to MANUAL D. 1. high

2. remains inAUTO Tuesday, July 13, 2010 Page 166 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 58 General Discussion With regards to a wet leg level transmitter on a reference leg leak the differential pressure between the reference leg and the variable leg goes to zero and the indicated level therefore fails high. With the Level Channel 1 transmitter input to the Level Select 1 failed to zero when the Level Channel 2 transmiiter input fails high (due to the reference leg leak), an alternate action is received (due to the deviation between level transmitter inputs) and the Level Master Controller swaps to manual. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible if the applicant confuses the operation of a dry leg level transmitter with awet leg level transmitter as this would be the correct response. Part 2 is correct. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible if the applicant confuses the operation of a dry leg level transmitter with a wet leg level transmitter as this would be the correct response. Part 2 is plausible if the applicant confuses the operation of the DCS Pressurizer Level Master Controller. If both transmitters failed low, it is plausible since there is no deviation between the two transmitters that an alternate action would not occur and the level controller would remain in auto. Additionally, if the applicant confuses the Level Master Controller with the NV-138 controller (which remains in auto on an Alternate Action) this answer is plausible. Answer C Discussion - -____________ CORRECT: See explanation above. iswer D Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is correct. Part 2 is plausible if the applicant confuses the operation of the DCS Pressurizer Level Master Controller. If the applicant does not recall that multiple transmitter failures where there is a deviation between the transmitters will result in an alternate action they would conclude that the Level Master Controller remains in auto. Additionally, if the applicant confuses the Level Master Controller with the NV-138 controller (which remains in auto on an Alternate Action) this answer is plausible. Basis for meeting the KA The KA is matched because a Pressurizer level reference leg failure has occurred and the applicant must determine the operational implication. In this case the Level Master Controller has swapped to MANUAL. Therefore, if actual Pressurizer level changes, the LCS will not respond to restore level. It is up to the Operator to restore level manually. - Basis for Hi Cog This is a higher cognitive level question because the applicant must associate multiple pieces of information both given and recalled from memory. The applicant is given that a transmitter input failure has occurred and must determine the direction in which the second transmitter fails and from both of those determine the effect on the LCS. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW flevelopment References udent References Provided rning Objective: i)PS-ILE-DCS#l2

References:

1) Lesson Plan OP.-MC-PS-ILE-DCS Tuesday, July 13, 2010 Page 167 of 294

____ FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination__ QUESTION 58 2558

2) DCS Control Builders Sheets t .PE028 AKI .01 Pressurizer (PZR) Level Control Malfunction

Knowledge of the operational implications of the following concepts as they apply to Pressurizer Level Control Malfunctions: (CFR 41.8 / 41.10/45.3) PZR reference leak abnormalities 401-9 Comments: RemarkslStatus 40 1-9 Comments: No comment. Resolution / Comments: N/A Tuesday, July 13,2010 Page 168 of 294

C) C) C 0 C m CI, 0 C Co m C, m Cl)

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 59 2559 APEO32 2.1.27 Loss of Source Range Nuclear Instrumentation

( PEO32 GENERIC i<nowledge of system purpose and/or function. (CFR: 41.7) Unit 1 is operating at 97% RTP when a Reactor Trip occurs. Given the following conditions: Channel Flux Level SUR SRN31 OCPS ODPM SR N32 0 CPS 0 DPM 1RN35 AMPS 10 1.1x1O -i/3DPM IR N36 11 AMPS 9.5x10 -1/3 DPM PRN41 12% PRN42 0% PRN43 0% PRN44 0% Which ONE (1) of the following statements describes why the Source Range Nuclear Instruments are NOT indicating? A. P-i 0 (Nuclear at Power) status light is LIT. B. P-6 (S/R Block Permissive) status light is LIT. C. P-i 0 (Nuclear at Power) status light is DARK. D. P-6 (SIR Block Permissive) status light is DARK. Tuesday, July 13, 2010 Page 169 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 59 2559 eneral Discussion 3 of 4 power range channels must be < 10% to auto-unblock SRNIs. Both intermediate range channels must be below P6 at lxlO-l0. When the P-6 permissive is DARK, the source range block permissive is removed and source range NIs will normally be energized. Based on one IR channel being greater than P-6, the P-6 permissive lit should be LIT and the SRs will not energize. Since 3/4 PR channels are less than P-b, the P-l0 permissive should be dark and therefore is NOT preventing the SRs from energizing. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant confuses the logic for the P-b permissive. The P-b Persmissive light being LIT would prevent the SRs from energizing. However, in this case since 3/4 PR channels are less than 10% the light should NOT be LIT. Answer B Discussion CORRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant confuses the logic for the P-b permissive or does not understand which state for the P-b Permissive light allows automatic re-energizing of the SRs. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant does not recall which state for the P-6 permissive light allows for automatic energizing of the SRs. Basis for meeting the KA For this question, the applicant is given indications that N41 has malfunctioned (indicating high). Additionally, they are given indications that N-36 is indicating high (potentially undercompensated) and preventing the P-6 permissive from clearing. This KA is matched since the operation of he P-6 permissive is a function of the system that in this case has resulted in a loss of both Source Range Nuclear Instruments. isis for Hi Cog ihis is a higher cognitive level question because it requires the applicant to analyze a given set of conditions and compare them to what the readings (recalled from memo) should be for this condition. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 2004 NRC Q36 (Bank 1236) Development References Ient References Provided Lesson Plan Objective: IC-ENB #12 IC-IPE #11

References:

1. OP-MC-IC-IPE Section 3.1.3

2. OP-MC-IC-ENB Sections 2.7 and 2.1.4 APEO32 2.1.27 Loss of Source Range Nuclear Instrumentation

APEO32 GENERIC Knowledge of system purpose and/or function. (CFR: 41.7) 401-9 Comments: RemarkslStatus 401-9 Comments: No comment. Resolution / Comments: N/A Tuesday, July 13, 2010 Page 170 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 59 Tuesday, July 13, 2010 Page 171 of 294

Question 59

References:

From Lesson Plan OP-MC-IC-IPE Section 3.1.3: Objective # 10 NC Pump Bus Under Frequency (214 busses = 56 Hz) this anticipatory loss of

coolant flow trip protects against DNB. The trip also trips open all four NC pump breakers to prevent electrical braking of the pump motors during frequency decay. A reduction in pump speed would reduce fly wheel inertia and pump coast down flow capability. This at-power trip protection is auto-blocked < 10% power (P-7) and is automatically reinstated > P-7. SG Lo-Lo Level (214 channels on 1/4 SGs = 17%) protects against a loss of heat

sink. This protection also causes an auto-start of the CA motor driven pumps (2/4 channels on 1/4 SGs) and the CA turbine driven Pump (2/4 channels on 2/4 SG5). Single Loop Loss of Flow (2/3 channels in 1/4 loops = 88%) protects against

DNB. This protection is auto-blocked <48% (P-8) and automatically reinstated> P-8. Two Loop Loss of Flow (213 channels in 214 loops = 88%) protects against DNB.

This protection is auto-blocked < 10% (P-7) and automatically reinstated> P-7. Safety Injection (any SI signal 1I2 Trains) initiates a reactor trip during LOCA

events. Turbine Trip (2/3 channels ASO <45psig, 414 stop valves closed) protects

against loss of integrity by preventing Pressurizer PORV5 from opening on turbine trip at high power. Objective #4, 10 General Warning (2/2 Trains) protects against a loss of both protection trains.

Anytime a General Warning is present on both SSPS trains a reactor trip will occur. General Warning is caused by: loose circuit board card; loss of voltage (AC or DC); SSPS train in Test; a Reactor Trip By-pass breaker in the Connected position and Closed; a Logic Ground Return fuse blown. 3.1.3 Protection Permissive Interlocks I Objective#11 I P-4 (Reactor Trip Breaker and Bypass Breaker Open for a given train) initiates:

Turbine Trip; Feedwater Isolation (coincident with low Tavg of 553 °F); Allows reset of SI signal after one minute time-out; Inputs to Steam Dump Control System for plant trip mode. P-6 (1/2 IR instruments> 10b0 amps) allows Manual Block of SR reactor trip. On

a power reduction, provides automatic reinstatement of SR high voltage and SR reactor trip when 2/2 IR channels < 10i amps. P-7 (2I4 PR instruments> 10% or 1/2 Turbine Impulse Pressures> 10%) Enables - (unblocks) the at power reactor trips: Pzr Hi-Level, Pzr Lo-Pressure, 2 Loop Loss of Flow, NCP UV, and NCP UF. The above trips are automatically blocked when below P-7, 3/4 PR < 10% and 2/2 Impulse Pressure < 10%.

Objective # 11 P-8 (214 PR instruments > 48% power) enables Single Loop Loss of Flow and

Reactor Trip upon Turbine Trip. P-IC (214 PR instruments> 10%) allows Manual Block of PR High Flux I Low

Setpoint reactor trip. Allows Manual block of lR High Flux Rod Stop (C-I) and Reactor Trip, blocks Manual reset of SR high voltage and SR reactor trip> P-IC. P-IC provides an input to P-7. Below P-IC (3/4 PR instruments < 10%) allows

Manual reset of SR High Voltage and Reactor trip. This is used if one IR channel does not decrease below P-6 to Auto energize the SR circuit. P-Il (2/3 Presurizer Pressure instruments < 1955 psig) allows Manual Block of

Lo-Pzr pressure SI (Auto instate> P-Il); allows Manual block of Lo Press Stm Line lsol (Auto instate> P-Il); Allows Manual block of motor driven CA pump Auto-start (Auto instate> P-I I); and initiates opening of Cold Leg Accumulator isolation valves when > P-I I. P-12 (2/4 Lo-Lo TAVG < 553°F) provides Auto-block of steam dumps preventing

excessive cooldown by the steam dumps. P-I3 (1/2 Impulse Pressure instruments> 10%) this turbine at power permissive

provides an input to P-7. P-14 (2/3 Hi-Hi level instruments on 1/4 SGs> 83%) actuates a Turbine Trip,

CFPT Trip and Feedwater Isolation. 3.1.4 Control Interlocks Objective # 12 C-I (1/2 IR channels amps > 20%) blocks Auto and Manual rod withdrawal.

C-2 (1/4 PR channels amps> 103%) blocks Auto and Manual rod withdrawal.

C-3 (2/4 AT channels within 2% of OTAT setpoint) blocks Auto and Manual rod

withdrawal plus actuates a turbine runback at 200%/mm for 2.3 seconds out of 30 seconds. C-4 (2/4 AT channels within 2% of OPAT setpoint) blocks Auto and Manual rod

withdrawal plus actuates a turbine runback at 200%/mm for 2.3 seconds out of 30 seconds. C-5 (I/I Impulse Pressure channels < 15%) blocks Auto rod withdrawal.

C-7A (Ill Impulse Pressure channel Ch II rate of change decrease > 5%lmin or a step change decrease> 10%) arms condenser dump valves on a load rejection.

From Lesson Plan OP-MC-IC-ENB Section 2.7: 2.7 Power Supplies NIS Channel I EKVA NIS Channel II EKVB NIS Channel Ill EKVC NIS Channel IV EKVD Wide Range Train A EKVA Wide Range Train B EKVD 3.0 SYSTEM OPERATION 3.1 Normal Operation 3.1.1 Operating Procedures The Excore Nuclear Instrumentation System provides the operator with neutron flux indication for all modes of operations. During each reactor startup, a healthy skepticism concerning the validity of power indications is warranted, particularly following a refueling outage. Changes in plant equipment or conditions, along with a strong desire to return the plant to full operation, may influence personnel to accept less than complete explanations for discrepant indications. For example, excessive electrical generation for the nuclear power indicated (a symptom of miscalibrated nuclear instruments) has been attributed to factors such as: cold circulating water temperature, expected efficiency improvements, and changes in core design or instrumentation.

From Lesson Plan OP-MC-IC-ENB Section 2.1.4: The Pulse Shaper shapes pulses into uniform square waves. The Pulse Driver matches impedance between the Pulse Shaper and the Log Pulse Integrator. The Log Pulse Integrator changes the pulse signal to a voltage output proportional to logarithm of pulse rate. The Level Amplifier amplifies the signal from the Log Pulse Integrator to drive bistables, indicators and other circuits. A Bistable Relay Driver provides the High Flux at Shutdown Alarm and the Containment Evacuation Alarm whenever the source range counts exceed the setpoint. Another Bistable Relay Driver provides the High Level Trip signal (1 cps). An isolation amplifier feeds the OAC, SUR Circuitry, Control Board Meter, and the NR 45 Chart Recorder. 2.1.3 Source Range Outputs Both Source Range channels read out on the Control Board with a range of iO° to 106 counts per second (cps). The Source Range level can be monitored on the NR-45 Control Room Chart Recorder. In addition to counts per second, Source Range Start up Rate (SUR) is indicated for each channel in decades per minute (-0.5 to 5.0 DPM). The High Flux at Shutdown alarm actuates when source range level reaches the setpoint of one half decade above normal shutdown counts. High Flux at Shutdown also actuates the Containment evacuation alarm inside the containment. A 5 second time delay precludes short duration spikes from actuating the Hi Flux at Shutdown and Containment Evacuation alarms. 2.1.4 Source Range Drawer Panel (Reference Figure 7.5). Objective # 8 Detector Volts Meter Indicates high voltage supplied to proportional counter detector.

Neutron Level Meter Scale 100 to 106 cps.

Instrument Power ON Lamp 118 volts AC Instrument power applied to drawer.

Control Power ON Lamp 118 volts AC Control Power applied to drawer.

Channel On Test Lamp Indicates the operation selector switch is in a position other

than NORMAL. Loss of Detector Volt Lamp Indicates high voltage to detector off or low.

Level Trip Lamp indicates neutron level greater than trip setpoint in Source Range.(1 5

cps) Level Trip Bypass Lamp On when Level Trip switch in Bypass for test or calibration.

HiQh Flux at Shutdown Lamp Neutron level greater than 1/2 decade above normal

shutdown level in Mode 6 and <5 times shutdown level in Modes 3,4&5. Bistable Trip Spare Lamp No function.

Instrument Power Fuses Overcurrent protection for power supply circuits. Instrument

power supplies the meters, circuit processing components, high voltage supply and detector power. This is true for the IR and PR drawers/circuits also. Control Power Fuses Overcurrent protection for control signal circuit transformers.

Control power supplies the lights on the drawer and 118 VAC to the bistable relay drivers to the plant relays. (High flux at shutdown alarm and SR high level trip). This is true for the lR and PR drawers/circuits also. NOTE (Reference Figure 7.21): If either instrument or control power fuses are removed, the bistables will trip. Level Trip Bypass will prevent bistable trip for Instrument Power fuses only. Objective # 10 Level Trip Switch Two position switch: Normal Switch Inactive; Bypass Enables

                    -                             -                            -

Operation Selector Switch for test and calibration; Provides AC signal to prevent Rx trip signal during testing. Operation Selector Switch Eight position switch enabled by Level Trip Switch to

Bypass position. Channel On Test lamp lights when not in Normal. Normal Switch - Inactive; Six Test Positions with Preset cps test values; Level Adjust Level Adjust

Potentiometer in circuit. Level Adjust Potentiometer Adjustable test signal into level amp. Enables adjustment

                               -                                     -

of the trip level of various bistables. Objective # 10 High Flux at Shutdown Switch Two position switch. Normal -allows circuit to provide

High Flux at Shutdown and Containment Evacuation alarm when setpoint is exceeded; Block-used during startup Blocks High Flux at Shutdown Alarm and

Containment Evacuation Alarm. 2.2 Intermediate Range 2.2.1 Intermediate Range Detectors Objective # 6 Reference Figure 7.6. Both intermediate range channels use compensated ion chambers to determine reactor power. These detectors are located just above the source range detectors in the same housing. The compensated ion chamber (CIC) uses two concentric Nitrogen gas filled, volumes: the outer is sensitive to both neutrons and gamma (boron lined); the inner sensitive only to gamma. As the two volumes are mounted concentrically in one unit, both are in essentially the same radiation field. By placing a negative potential on the inner lead, the gamma signal generated in the inner volume is made to compensate or cancel out the gamma signal generated in the outer volume. Since the two volumes can not be manufactured exactly the same size, the high voltage to the center electrode is variable to adjust the sensitivity of the inner volume. Operating in the recombination region, a change in inner

volume detector voltage will vary the gamma current for a given flux level. The outer volume operates in the ion chamber region where all the ion pairs are collected.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2004 CNS SRO NRC Examination QUESTION 36 fl36 APEO32 AA2.05 Loss of Source Range Nuclear Instrumentation

Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: (CFR: 43.5 / 45.13) Nature of abnormality, from rapid survey of control room data Unit I is operating at 97% power when a reactor trip occurred. Given the following conditions: Channel Flux Level SUR PRN44 0% PRN43 11% PRN42 0% PRN4I 12% IR N36 11 9x10 -1/3 DPM IR N35 11 5x10 -1/3 DPM SR N32 0 CPS 0 DPM SR N31 0 CPS 0 DPM Which one of the following statements correctly describes why the source range instruments are not indicating? A. P-6 (S/R Block Permissive) status light is DARK. B. Loss of power to bus I ERPD. C. P-lU (Nuclear at Power) status light is LIT. D. Loss of power to bus 1ERPB. Tuesday, July 13, 2010 Page 72 of 203

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2004 CNS SRO NRC Examination QUESTION 36 l236 General Discussion Bank Question: 1146 3 of 4 power range channels must be < 10% to auto-unblock SR NIs. Both intermediate range channels are below P6 at lx 10-10. When the P-6 permissive is DARK, the source range block permissive is removed and source range NIs will normally be energized. The P-b permissive prevents this normal operation. Answer A Discussion Plausible: If the candidate thinks that intermediate range NIs are not below P-6 or does not recognize that 3 of 4 power range channels must be < 10%. Answer B Discussion Incorrect: Loss of 1ERPD does not affect source range. Plausible: Loss of 1ERPD would cause N44 to read zero power. If the candidate thinks that loss of IERPD can deenergize P-b, and reverses the effect ofP-10 on loss of power. Answer C Discussion rCoct 3 of 4 power range channels must be < 10% to auto-unblock SR NIs. NI-43 and NI-4 1 are still above NOT P- 10. Answer D Discussion Incorrect: Loss of 1ERPB would cause only N42 to read zero. Plausible: If the candidate thinks that loss of 1ERPB can deenergize P-b, and reverses the effect of P-b on loss of power. Basis for meeting the KA rBasis for Hi Cog Basis for SRO only E Job Level Cognitive Level QuestionType Question Source

RO Comprehension BANK Development References ient References Provided Lesson Plan Objective: IC-ENB SEQ 6

References:

1. OP-CN-IC-ENB page 10, 25, 41

2. OP-CN-IC-IPX page 37 APEO32 AA2.05 Loss of Source Range Nuclear Instrumentation

Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: (CFR: 43.5 /45.13) Nature of abnormality, from rapid survey of control room data O1-9 Comments: _____________ rnarksIStatus - -- Tuesday, July 13, 2010 Page 73 of 203

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 60 APEO33 AA1.03 Loss of Intermediate Range Nuclear Instrumentation

bility to operate and / or monitor the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: (CFR 41.7/45.5 / p5.6) Manual restoration of power Given the following conditions on Unit 1:

Unit is currently at 35% RTP
A unit shutdown is in progress
Intermediate Range Channel N35 fails
N35 Level Trip Bypass switch has been placed in BYPASS in accordance with AP-16 (Malfunction of Nuclear Instrumentation)
N35 Instrument Power fuses and Control Power fuses have been removed for troubles hooting Which ONE (1) of the following describes the actions required to prevent a Reactor Trip and the MINIMUM power level at which those actions must be performed if the unit shutdown is continued?

A. N35 Control Power fuses ONLY must be installed or a Reactor Trip will occur when power decreases to less than 10% RTP. B. N35 Control Power fuses ONLY must be installed or a Reactor Trip will occur when power decreases to less than 25% RTP. C. N35 Control Power fuses AND Instrument Power fuses must be installed or a Reactor Trip will occur when power decreases to less than 10% RTP. D. N35 Control Power fuses AND Instrument Power fuses must be installed or a Reactor Trip will occur when power decreases to less than 25% RTP. Tuesday, July 13, 2010 Page 172 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 60 2560 A General Discussion When the Control Power and Instrument Power fuses are removed, the N35 bistables will trip (i.e. JR High Flux Trip). Unlike a Power Range channel where removing only the Control Power fuses will cause the bistables to trip, in the case of the JR channel, removing EITHER the Control Power or Instrument Power fuses will cause the bistables to trip. In this particular case, even though the bistables are tripped, a Reactor Trip does not occur. The JR Hi Flux trip had been previously blocked when the P-b permissive was met (> 10% Power). If the unit shutdown was continued and power decreased to less than 10% RTP, the reactor would trip on IR Hi Flux unless BOTH the Control Power and Instrument Power fuses are installed. Alternatively, the Control Power fuses could be installed AND the Level Trip Bypass switch placed in BYPASS to prevent a Reactor Trip. Answer A Discussion CORRECT: See explanation above. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because the Control Power Fuses must be installed. Installing the fuses prior to decreasing to less than [25% power is plausible since this is the equivalent power at which the JR Hi Flux Trip occurs. - Answer C Discussion ._________ INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because both set of fuses would have to be installed prior to decreasing power to less than 10% power if the Level Trip Bypass Switch had not been placed in bypass as directed by procedure. Answer 0 Discussion INCORRECT: See explanation above. AUSIBLE: This answer is plausible because both sets of fuses would have to be installed to prevent a trip if the Level Trip Bypass switch had not been placed in bypass as directed by AP-16. Installing fuses prior to 25% power is plausible since this is the equivalent power at which the JR Hi Flux Trip occurs. Basis for meeting the KA This KA is matched because the applicant will demonstrate the ability to operate the IR Nuclear Instrumentation by demonstating a knowledge Instrument Power and Control Power fuses and when power must be restored to the channel_during a unit shutdown. Basis for Hi Cog This is a higher cognitive level question because it requires the applicant to associate multiple pieces of information. First, the applicant must recall from memory the effect of removing the control power and instrument power fuses from the JR channel. The applicant must then compare [that information to the conditions given in the stem of question to determine the effect of continuing the shutdown with the fuses removed and what actions must be taken to continue the shutdown without causing a Reactor Trip. Since this question requires more than one mental step to arrive at the correct answer, this is a higher cognitive level question. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension NEW Development References Student References Provided Learning Objective IC-ENB #8

References:

1. Lesson Plan OP-MC-IC-ENB Section 2.2.5 Lesson Plan OP-MC-IC-IPE Figure 7.7 -_____________

APEO33 AA1.03 Loss of Intermediate Range Nuclear Instrumentation

Ability to operate and/or monitor the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: (CFR 41.7 / 45.5 / 45.6) Manual restoration of power Tuesday, July 13, 2010 Page 173 of 294

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2010 MNS SRO NRC Examination QUESTION 60 2560 1-9 Comments: marksIStatus 401-9 Comments: The way these are written, if B is correct then A is correct. If D is correct, C would be correct too. Qualifiers need to be added to eliminate the subsets. This Q is U because of two invalid responses. Resolution I Comments: We had lengthy discussions about subset issues when we wrote this question and specifically wrote the question to include a Reactor Trip will occur when because we felt it was a qualifier that eliminated the subset issue. However, re-wrote the stem to add another qualifier there too. See attached file for two versions of proposed revision. Tuesday, July 13, 2010 Page 174 of 294

Question 60

References:

From Lesson Plan OP-MC-IC-ENB Section 2.2.5: Bistable Relay Drivers provide the P-6 Permissive, the Low Power Rod Stop and the Reactor Trip whenever the intermediate range amps exceed the setpoint. An isolation amplifier feeds the OAC, SUR Circuitry, Control Board Meter, and the NR-45 Chart Recorder. 2.2.4 Intermediate Range Outputs Both Intermediate range channels read out on the Control Board with a range of 10.11 to i0 amps. The Intermediate Range level can be monitored on the NR-45 Control Room Chart Recorder. In addition to counts per second, Intermediate Range Start-Up Rate (SUR) is indicated for each channel in decades per minute (-0.5 to 5.0 DPM). The P-6 Source Range Block Permissive actuates when 1-out-of-2 (1/2) IR channels exceeds 10b0 amps. The Low Power Rod Stop prevents outward motion of the rods in Auto and Manual when 1/2 lR channels exceeds amps equivalent to 20% reactor power. The Low Power Reactor Trip protects the core from startup accidents when 112 IR channels exceeds amps equivalent to 25% reactor power. 2.2.5 Intermediate Range Drawer Panel (Reference Figure 7.9). E Objective # 8 I Ampere Neutron Level Meter Indicates current output of detector in amps with a range

of eight decades (10-11 to 1 0 amps) Instrument Power ON Lamp 118 volt AC instrument power is applied to drawer.

Control Power ON Lamp 118 volt AC control power is applied to driver assembly

control circuits. Channel On Test Lamp Indicates Operation Selector switch is in a position other than

NORMAL. Level Trip Bypass Lamp Indicates Level Trip switch in BYPASS position.

HiQh Level Trip Lamp ON when neutron flux in IR exceeds current equivalent to 25%

full power. (Approximately 5.5 x i0 amp). Hicih Level Rod Stop lamp ON when IR current equivalent to 20% full power.

Power Above Permissive P-6 Lamp Lights when IR reaches 10-10 amps. Allows

blocking Source Range Instruments. Loss of Detector Volt Lamp Indicates low or loss of high voltage to detector.

Loss of Comp. Volt Lamp Indicates loss of compensating voltage to detector.

AC Inst. Power Fuses Overcurrent protection for instrument power.

AC Control Power Fuses Overcurrent protection for control power.

NOTE (Reference Figure 7.21): If either instrument or control power fuses are removed, the bistables will trip. Level Trip Bypass will prevent bistable trip for Instrument Power fuses only.

From Lesson Plan OP-MC-IC-IPE Figure 7.7: 7.7 Protection Permissive Interlocks (12/17/99) INTERLOCKS LOGIC FUNCTION P-i0 2/4 P.R.> 10% FP On increasing power P-10 allows manual block of the Intermediate Range trip and rod stop (C-i). Allows block of the Power Range High Flux Low Setpoint trip and prevents the Source Range instruments from being Manually energized. (Will automatically de-energ ize both source range detectors if not previously de energized at P-6.) Also provides an input to P-7. On decreasing power, the Intermediate Range trip and the Power Range trip are automatically reactivated, allows manual reset of SR High Voltage block if one IR channel does not decrease below P-6 to auto energize the SR circuit. P-lI 2/3 Pzr Press < 1955 On decreasing pressure (<1955 #) P-il allows manual block of Low Pzr Pressure Safety Injection, Lo Press Stm Line Isol and CA Pump Auto start. Enables High Steam Rate Main Steam Isolation. P-12 2/4 Lo-Lo Tave < 553 °F Blocks steam dumps P-13 1/2 Impulse Press> 10% Input to P-7 P-14 2/3 Level on 1/4 SIG Hi-Hi

Turbine Trip Level > 83%
FWPT Trip
Feedwater Isolation

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 61 APEO59 AK1.0l Accidental Liquid Radioactive-Waste Release

Thowledge of the operational implications of the following concepts as they apply to Accidental Liquid Radwaste Release: (CFR 41.8 /41.10 /

,5.3) ypes of radiation, their units of intensity and the location of the sources of radiation in a nuclear power plant An NEC reports a leak in the 1A NV pump room. Water from the leak is spraying into the air and is also collecting on the floor.

Radiation Protection (RP) is notified and determines the water is contaminated. RP reports that the radiation from the leak is exclusively a skin dose and absorption concern The radiation emitted from the contamination is predominately (1) This event results in a Trip 2 alarm on 1 EMF-41 (Aux Building Ventilation). The Auxiliary Building Ventilation (2) Which ONE (1) of the following completes the statements above? A. 1. Alpha

2. supply and exhaust fans will trip B. 1. Beta

2. supply and exhaust fans will trip C. 1. Alpha

2. filter train will be placed in service D. 1. Beta

2. filter train will be placed in service Tuesday, July 13, 2010 Page 175 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 61 25611 General Discussion The radiation that poses a skin dose concern is Beta. Alpha is shielded by layers of clothing or paper and poses no skin dose threat. On a Trip 2 alarm on EMF-41, the Auxiliary Building Ventilation system filter train will be placed in service (un-bypassed). Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible if the applicant does not recall the concern with Alpha and Beta radiation. Part 2 is plausible if the applicant does not recall the automatic actions caused by EMF-41. It is plausible to believe that the supply and exhaust fans would trip to prevent the spread of airborne radiation. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is correct. Part 2 is plausible if the applicant does not recall the automatic actions caused by EMF-41. It is plausible to believe that the supply and exhaust fans would trip to prevent the spread of airborne radiation. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible if the applicant does not recall the concern with Alpha and Beta radiation. Part 2 is correct. Answer D Discussion ORRECT: See explanation above. asis for meeting the KA The KA is matched because and accidental liquid radwaste release has occurred and the applicant must know the type of radiation causing the concern based on the description and the operational concern based on automatic actions that will occur as a result of the release. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK 2006 NRC Q22 (Bank 628) Development References Student References Provided Learning Objectives:

1) RAD-RP#6

2) WE-EMF#2,3

References:

1) NANTEL Generic Radiation Working Training Lesson Plan

2) Lesson Plan OP-MC-WE-EMF Section 2.1.6 APEO59 AK1.01 Accidental Liquid Radioactive-Waste Release

Knowledge of the operational implications of the following concepts as they apply to Accidental Liquid Radwaste Release: (CFR41.8/4l.l0 I 45.3) ( s of radiation, their units of intensity and the location of the sources of radiation in a nuclear power plant 401-9 Comments: RemarkslStatus 401-9 Comments: Tuesday, July 13, 2010 Page 176 of 294

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2010 MNS SRO NRC Examination QUESTION 61 1 am not convinced that the stem excludes alpha The distractor analysis is not clear on it either. The reference states that it is an internal hazard. The stem references an airborne concern. Add something to the stem to eliminate C as a potential correct answer. Resolution I Comments: Changed stem of question to read The radiation emitted from the water on the floor is primarily . This will eliminate alpha as a potential answer because it eliminates the airborne contamination. The only reason the airborne contamination is included is to give plausibility to the EMF-41 alarm. See attached file for_proposed revision. Tuesday, July 13, 2010 Page 177 of 294

Question 61

References:

From NANTEL Generic Radiation Worker Training Lesson Plan: NANTeL Generic Radiation Worker Training Lesson Pan 11

exposure hazard (for example, whole-body, skin, eyes)
major sources
Penetrating ability: least penetrating ability of the four types; travels approximately one inch in the air.
Shielded by: a piece of paper, lightweight clothing, or outer layer of skin
Type hazard: internal hazard can result in high dose to sensitive organ
Major source: nuclear fueli
Penetrating ability: travels a few feet in the air
Shielded by: lightweight plastic or aiuminuF
Type hazard: eyes and skin are most susceptible to beta radiation. It can be ar nternal hazard.

   . Major source: most beta particles come from activated corrosion and fission products.

L Additional information: personnel must work close to a beta source to receive uch do Gamma

Penetrating ability: very high. Penetrates the whole body.
Shielded by: very dense material; usually lead, steel, water, and concrete.
Type hazard: whole-body dose hazard.
Major source: fission, fission products, and activation products in the primary system piping.
Additional information: has no mass or electrical charge it is pure energy.

Neutron

Penetrating ability: very high.
Shielded by: water, paraffin, or concrete.
Type hazard: whole-body dose hazard.
Major source: mainly a problem only near the reactor when it is operating.
Additional information: Neutrons are freed from the nucleus by decay or fission and have no electrical charge.

Where You Will Find Alpha Radiation Alpha radiation is primarily found in or near the fuel assemblies. Remember that alpha radiation cant even penetrate paper. Since the fuel is contained in metal rods, the main hazard from alpha radiation exists only if the fuel assembly develops leaks. Where You Will Find Neutron Radiation Neutron radiation is normally a concern only when the reactor is running. This is

From Lesson Plan OP-MC-WE-EMF Section 2.1.6: 2.1.6 Auxiliary Building Ventilation Monitor The Auxiliary Building is monitored by OEMF 41 Aux Building Ventilation.

Objective # 2, 5 EMF-41 uses a scanner capable of monitoring 12 points within the Auxiliary Building ventilation ducts. These points are located to provide maximum coverage of Auxiliary Building rooms. (refer to Drawing 7.2 and 7.3) NOTE: Sample point 6 has been deleted, so only 11 points are currently monitored. A timed sample system is used to control the solenoid valves (refer to Drawing 7.4). Each sample point takes about 2.5 minutes, 1.5 minutes to purge and 1 minute to sample. Thus, each point will be sampled twice per hour. The flow rate for each sample line is 1 scfm. This 1 scfm from the sampled line is routed through the detector. A SCAN/STOP switch is provided to control EMF 41 operation mode: refer to Drawing 7.4)

Scan Mode provides automatic sequential sampling of 11 Aux. Bldg areas. PT/1/A/4600/038

requires the toggle switch to be in the scan position.

Stop - provides continuous sampling of one area.

A ready light illuminates while EMF is sampling and off while purging. A STEP switch allows manual

selection of desired sample point. This option is available in the SCAN mode only. A point window provides an LED readout that displays selected sample point. When the scanner is selected to a single point, remote readout to the OAC and P1 database is disabled. Only the local Control Room module readout is available. Objective # 2, 3 On a Trip 2 high radiation alarm, Aux. Building Ventilation will be passed through filter units ABFU-1 and ABFU-2 (filter bypass will be terminated). The following dampers will open:

1ABF-D-4A 2ABF-D-4A
1ABF-D-4B 2ABF-D-4B
1ABF-D-5A 2ABF-D-5A
1ABF-D-5B 2ABF-D-58 The following dampers will close:
1ABF-D-3
2ABF-D-3
1ABF-D-6
2ABF-D-6

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2006 CNS SRO NRC Examination QUESTION 22 628 N/A - N/A Never Assigned to a K/A An NLO is performing a leakage check to determine identified leakage for use during the NC Leakage Test. While completing the check he inadvertently spills 1 gallon of liquid on the floor. Radiation Protection (RP) is notified and determines the water is contaminated. RP reports the radiation is primarily a skin dose and absorption concern. Which one of the following correctly describes the type of radiation which is emitted from the fluid and where this radioactive material comes from? A. Alpha Tritium due to activation of hydrogen in the reactor coolant B. Beta Tritium due to activation of hydrogen in the reactor coolant C. Alpha Cobalt due to activation of corrosion or wear products D. Beta Cobalt due to activation of corrosion or wear products Tuesday, July 13, 2010 Page 44 of 203

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2006 CNS SRO NRC Examination QUESTION 22 628 General Discussion Answer A Discussion Incorrect: Plausible: Second part of answer is correct but alpha is not a skin dose concern and is not normally emitted by liquids. Answer B Discussion Answer C Discussion Incorrect Plausible: Alpha is not a skin dose concern and activated corrosion or wear products cannot be absorbed thru the skin. Answer D Discussion D. El Incorrect: LiPlausible: Beta is correct but activated corrosion or wear products cannot be absorbed thru the skin. Basis for meeting the KA Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK Development References Student References Provided GET Manual N/A - N/A Never Assigned to a K/A 401-9 Comments: Remarks/Status Tuesday, July 13,2010 Page 45 of 203

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 62 APEO68 AA 1.01 - Control Room Evacuation ( \bility to operate and/or monitor the following as they apply to the Control Room Evacuation: (CFR 4 1.7/45.5 /45.6) i/G atmospheric relief valve Given the following plant conditions:

The Control Room has been evacuated due to a chlorine gas leak
AP-17 (Loss of Control Room) has been implemented on both units
An Operator has been dispatched to control 1A & 1D SG PORVs using AP-17, Enclosure 7 (Manual Control of PORVs)

The Operator dispatched to perform Enclosure 7 will control the SG PORVs using manual loaders located in the (1) To establish control of the SG PORVs the

Operator must (2) Which ONE (1) of the following completes the statements above? A. 1. Exterior Doghouse

2. open the VI supply from the local manual loader ONLY B. 1. Interior Doghouse

2. open the VI supply from the local manual loader ONLY C. 1. Exterior Doghouse

2. open the VI supply from the local manual loaderAND close the VI supply from the Control Room manual loader D. 1. Interior Doghouse

2. open the VI supply from the local manual loader AND close the VI supply from the Control Room manual loader Tuesday, July 13, 2010 Page 178 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 62 2562 C jeneral Discussion - Local Manual Operation

This capability is provided for A & D Steam line PORVs only. The control stations are located in the Exterior Doghouse near their respective PORVs. The Operator must isolate VI supply from Control Room manual loader and open supply from local manual loader. Operator may now open/close PORV using the local manual loader (valves will no longer operate automatically in this mode). Answer A Discussion -_______________ NCORRECT: See explanation above. PLAUSIBLE: Part I is correct. Part 2 is plausible if the applicant does not understand that the PORV can still be controlled by the Control Room manual loader if VI from that loader is not isolated. Answer B Discussion INCORRECI: See explanation above. PLAUSIBLE: Part 1 regarding the Interior Doghouse is plausible if the applicant does not recall the location of the manual loader station. Part 2 is plausible if the applicant does not understand that the PORV can still be controlled by the Control Room manual loader if VI from that loader is not isolated. Answer C Discussion E:See_explanation above. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 regarding the Interior Doghouse if plausible if the applicant does not recall the location of the manual loader station. art 2 is correct. asis for meeting the KA The KA is matched because a Control Room Evacuation has occurred and the applicant demonstrates an ability to operate the SO PORVs (SG atmospheric relief valve) by demonstrating a knowledge of the location of the local control stations and the requirements to establish local control of the PORVs. Basis for Hi Cog -.- . - Basis for SRO only E L Job Level Cognitive Level QuestionType Question Source -_____ RO Memory T NEW Development References Student References Provided Learning Objective:

1) STM-SM #9

References:

1) Lesson Plan OP-MC-STM-SM Section 2.3.2

2) AP/17 Enclosure 7 .__________________

APEO68 AA1.01 Control Room Evacuation

Ability to operate and / or monitor the following as they apply to the Control Room Evacuation: (CFR 41.7 / 45.5 / 45.6) Sf0 atmospheric relief valve 401-9 Comments: RemarkslStatus 40 1-9 Comments: I believe the interior doghouse will be readily eliminated due to Tuesday, July 13, 2010 Page 179 of 294

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2010 MNS SRO NRC Examination QUESTION 62 common knowledge. Replace BI and DI. This Q is U due to Bl and Dl beingNP Resolution / Comments: Believe that if we replace RI and Dl with anything else it will be less plausible than it is right now. Open to_suggestions. Tuesday, July 13, 2010 Page 180 of 294

Question 62

References:

From Lesson Plan OP-MC-STM-SP4 Section 2.3.2: Objective #9 2.3.2 PORVs Modes of operation (Refer to STM-SM-2) Automatic The manual loader is normally in auto set at 100% open. Steam line pressure increases to open setting (1125 psig). Limit switch opens 3-way solenoid valve to admit Vito open the PORV. The VI pressure must pass through the manual loader. The setting of the manual loader (normally set at 100% open) determines how far the PORV will open whenlif the setpoint is reached. When steam line pressure decreases to 1092 psig the 3-way solenoid valve repositions to vent the air pressure on the valve and blocks air pressure to the valve positioner. Manual Operator selects manual to admit VI to positioner. PORV will open to manual loader valve position. Prior to selecting manual, operator should close all PORVs manual loaders to prevent inadvertent operation of steam line PORVs. Local - Manual Operation This capability is provided for A & D Steam line PORVs only. The control stations are located in the Exterior Doghouse near their respective PORV. The Operator must isolate VI supply from Control Room manual loader and open supply from local manual loader. Operator may now openlclose PORV using the local manual loader (valves will no longer operate automatically in this mode). All the PORVs have a local valve operator extension attached to the valve actuator for true Manual operation. If the Local-Manual control station operation does not work, then the operator can manually operate the PORV. All S1G PORV manual operators are reverse acting (Clockwise to open), 2B AND 2C SIG PORVs are the exceptions which operate per the normal convention (Counter-clockwise to open).

From APII7 Enclosure 7: MNS LOSS OF CONTROL ROOM PAGE NO. AP/21A/5500/1 34 of 41 Enclosure 7 Page 1 of 2 Rev. 19

UNIT 2 Manual Operation of PORVs A::T:ON/EXPECTED RESPONSE RESPONSE NDT OBTAINED Establish communication from doghouses to SRO at Aux Shutdown panel. NOTE A Main Steam Isolation signal or Loss of VI will prevent operation of PORVs from manual loaders.

2. Operate valves 2SV-I9AB (2A Main Operate the following valves PER Steam Line PORV) and 2SV-IAB (2D instructions near valves:

Main Steam Line PORV) (exterior doghouse) using manual loaders as

2SV-19AB (2A Main Steam Line PORV) follows:

                                                              . 2SV-1AB (2D Main Steam Line PORV).

a. Ensure the following controller knobs are in the full counter clockwise position:

Manual loader 2SMML5521 (2A SM PORV (2SV-19) Local Manual Loader)

          .      Manual loader 2SMML5491 (2D SM PORV (2SV-1) Local Manual Loader).

b. Ensure the following valves are open:

              . A-i (2A SIG Local Manual Loader Input Isol)

              . D-1 (2D SIG Local Manual Loader Input Isol).

c. Close the following valves:

A-2 (2A SIG Control Room Manual Loader Output Isol)

              . D-2 (2D S/G Control Room Manual Loader Output Isol).

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 63 2563 APE069 2.4.50 Loss of Containment Integrity

PE069 GENERIC ibility to verifi system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10/43.5/45.3) Given the following conditions on Unit 1:

The unit was operating at 100% RTP with a VQ release in progress
A Rx Trip was manually initiated due to the 1A SIG FRV failing closed
The resulting transient resulted in a tube rupture on the 1A SIG
The crew has manually initiated Safety Injection
Both trains Cont Vent Isol Reset lights on IMC-11 are LIT
1VQ-1A (U-i Cont Air Release Inside Isol) indicates OPEN
No AUTO SI setpoints have been exceeded
1EMF 38, 39 & 40 readings have remained less than Trip 2 values Based on these conditions, the Containment Ventilation Isolation Reset Lights should be (1) AND the Operators shall (2)

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

2. close 1VQ-1A B. 1. DARK

2. verify 1VQ-1A remains OPEN C. i. LIT

2. close IVQ-IA D. i. LIT

2. verify 1VQ-iA remains OPEN Tuesday, July 13, 2010 Page 181 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 63 2563

  ..eneraI Discussion The question is providing a set of conditions where a Manual SI has been initiated but there has been no Auto SI or Phase B setpoints exceeded.

A Containment Isolation signal (Sh) is generated by anyone of the 4 following signals: Safety Injection, Manual Phase A, Manual Phase B, and a Trip 2 on EMF-38, 39, or 40. Since a Ss signal has been generated there should have been a Sh signal as well and the Sh reset lights should be dark. 1VQ-l closes on a Sh signal and should be closed in the scenario given in the stem of the question. Answer A Discussion ECORRECT See explanation above. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: First part of the answer is correct. The second part is plausible if the applicant confuses the signals which will close this valve. One of the initiating signals which will generate and Sh signal is a Manual Phase B. Answer C Discussion -________________ INCORRECT: See explanation above. PLAUSIBLE: The first part is plausible if the applicant confuses a manual SI with the other manual isolation signals which will generate an Sh. The list of signals only lists a SI signal and doesnt differentiate between manual or Auto. The second part of the answer is correct. Answer D Discussion See explanation above. PLAUSIBLE: The first part is plausible if the applicant confuses a manual SI with the other manual isolation signals which will generate an Sh. The list of signals only lists a SI signal and doesnt differentiate between manual or Auto. e second part is plausible if the applicant confuses the signals which will close this valve. One of the initiating signals which will generate an Sh signal is a Manual Phase B. Basis for meeting the KA The K/A is matched because the applicant must demonstrate the ability to verify that the containment ventilation isolation Train A and Train B reset lights are indicating correctly for given plant conditions and what control board actions is to be taken in response. The scenario given represents a loss of containment integrity. Basis for Hi Cog This is a hi cog question because it involves a level of analysis of given situation, apply system knowledge and solve a problem. Basis for SRO only

 -  Job Level               Cognitive Level           QuestionType                                    Question Source RO                  Comprehension                 NEW velopment References                                                                            Student References Provided Lesson Plan OP-MC-ECC-ISE Page 19 (Rev 31)

Lesson Plan OP-MC-CNT-VQ Page 13 (Rev 18) Lesson Plan OP-MC-CNT-VQ Page 31 (Rev 18) OP-MC-ECC-ISE Obj: 5 OP-MC-CNT-VQ Obj: 5 PE069 2.4.50 Loss of Containment Integrity

( E069 GENERIC oility to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10 / 43.5 /45.3) Tuesday, July 13, 2010 Page 182 of 294

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2010 MNS SRO NRC Examination QUESTION 63 2563

,4O1-9 Comments:                    Remarks!Status

( 40 1-9 Comments: Suggestion: Change A2 and D2 to verify 1VQ-1A remains open to make these a little more plausible. Resolution / Comments: Revised question per Lead Examiners recommendation. Also, changed first should to will and second should to shall in the stem of the question per Lead Examiners General Comments. See attached file for revised question. -_____ Tuesday, July 13, 2010 Page 183 of 294

Question 63

References:

OP-MC-ECC-ISE Obj: 5 OBJECTIVES NNLLL OBJECTIVE L L P P 0 OORSR RO0 1 State the purpose of the Engineered Safeguards System. X X X X 2 Explain the need and reasoning behind the redundancy X X X X requirements for two trains of safety related systems. 3 State how the operator would be aware if more than one X X X protection cabinet door was opened simultaneously. 4 Define the following terms: X X X X Ss St, Sp, SH 5 List the conditions that will initiate the following: X X

        . Safety Injection (Se)
        . Phase A Isolation (Si)
        . Containment Spray/Phase B Isolation (Sp)
        . Containment Ventilation Isolation (SH)
       . Main Steam Isolation (MSI)
       . Main Feedwater Isolation (FWI)
       . VE (Annulus Ventilation) System Start
       . H  2 Skimmer and Air Return Fan Start (VX) 6  List all Safety Injection (Se) actuation signals, setpoints, logic, X X  X X X and the type of accident each signal provides protection for.

7 List the pumps that automatically start following a safety X X X X X injection actuation. 8 State which Safety Injection (S) signal can be blocked. X X X X X 9 Explain the reason for blocking a Safety Injection (Se) signal. X X X X X 10 List the interlock and parameter setpoint that allows blocking X X X X X Safety Injection (Se). 11 Describe the operator action needed to block Safety Injection. X X X 12 List the conditions that allow RESET of Safety Injection. X X X

OP-MC-CNT-VQ Obj: 5 OBJECTIVES NNLLL OBJECTIVE L L P P 0 OORSR ROO 1 State the purpose of the Containment Air Release and X X X X Add ition System ( VQ). 2 Draw the VQ System per Drawing 7.1 and/or describe the X X X X flow path for addition and release. 3 Describe the normal operating conditions which will result in X X X X containment pressure increases or decreases ( excluding primary and secondary leaks). 4 State why air removal is from lower containment and air X X X X addition is to upper containment. 5 State which signal will isolate the VQ System. X X X X X 6 Describe the Control Room instrumentation and controls X X X associated with the VQ System. 7 Given a limit and/or precaution associated with the VQ X X X X X System, discuss its basis and when it applies. 8 State the differences between a VQ release with the totalizer X X X operable versus a release with the totalizer inoperable.

From Lesson Plan OP-MC-ECC-ISE Page 19 (Rev 31) Objective # 5 Containment Ventilation Isolation (SHI

Safety Injection (Si)
Manual Phase A (Se)
Manual Phase B
Trip 2 on EMF-38, 39, or 40 Main Steam Isolation IMSI)
Hi Hi Containment Pressure (Sr)
Low Steamline Pressure
High Steam/me Pressure rate of decrease (below P-li with Lo Press Stm Line Isol blocked)
Manual Main Feedwater Isolation (FWI)
Safety Injection (S)
Reactor Trip and Low T-avg
High High S/G Level
Manual VE (Annulus Ventilation) System Start
Hi Hi Containment Pressure (S,)
Manual H2 Skimmer and Air Return Fan Start (VX)

Hi Hi Containment Pressure (Sr) CPCS 10 minute time delay

From Lesson Plan OP-MC-CNT-VQ Page 13 (Rev 18) 2.0 COMPONENT DESCRIPTION 2.1 PAC Filters The VQ filters are PAC filters. The Particulate section removes large particles. The Absolute section removes small particulate while the Charcoal section removes iodine. A local pressure gauge to indicate D/P across the filters is checked on the Unit #2 Auxiliary Building rounds sheets for the 767 elevation. A satisfactory reading is less than or equal to 3.4 inches WC. When a filter DIP reaches 6.0 inches WC, the other filter is placed in service. Normally only one filter is in service at a time. 2.2 Valves Objective # 5 VQ-IA (Inside) and VQ-2B (outside) Containment Air Release Isolation valves are air operated diaphragm valves operated from the Control Room on MC-1 1 to align the release flow path. This valve will fail closed on a loss of air. These valves will auto close on an SH (Containment Ventilation Isolation) signal. Note that an automatic or manual safety injection will generate a containment ventilation isolation signal. VQ-3 Containment Air Addition Inlet from Auxiliary Building Isolation Valve is a piston operated gate valve. This valve is manually opened with VQ-6A and VQ5B on a -0.20 psig Containment pressure signal to allow air to be drawn in from the Auxiliary Building. VQ-3 fails closed on a loss of air. Objective #6 VQ-4 Containment Air Release to Unit Vent is an air operated gate valve which has a manually operated loader with open/close indication located on MC-9 ( refer to Drawing 7.3). This valve is adjusted to control the release rate. This valve will fail closed on a loss of air. VQ-5B( outside) and VQ-6A( inside) Containment Air Addition Isolation valves are air operated diaphragm valves operated from the Control Room on MC-1 I to align the flow path for air addition. This valve will fail closed on a loss of air. These valves will auto close on an SH (Containment Ventilation Isolation) signal. Note that an automatic or manual safety injection will generate a containment ventilation isolation signal.

From Lesson Plan OP-MC-CNT-VQ Page 31 (Rev 18) UPPER CONTAI NM ENT 4 H 5 6A LOWER CONTAI NMENT 1A AUX BLDG

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 64 EPEO74 EA2.Ol Inadequate Core Cooling

bility to determine or interpret the following as they apply to a Inadequate Core Cooling (CFR 43.5 / 45.13)

.,ubcooling margin Which ONE of the following will generate a LOSS OF SUBCOOLING (AD2-D5) annunciator in the Control Room?

A. 2°F subcooling on Loop A TH B. 0°F subcooling on Loop B TH C. 2°F subcooling on Loop C TH D. 0°F subcooling on Loop D TH Tuesday, July 13, 2010 Page 184 of 294

0 FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 64 2564

    ,neral Discussion The Loss of Subcooling (AD2 I D5) annunciator alarms if subicooling decreases to 0°F subcooling on Loop D Thot.

Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausibe because the SUBCOOLING MARGIN ALERT (AD2 I D4) annunciator alarms at 2°F subcooling and it is sensed from Thot, just a different ioop. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because 0°F subcooling is correct and it is sensed from Thot, just a different loop. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausibe because the SUB COOLING MARGIN ALERT (AD2 / D4) annunciator alarms at 2°F subcooling and it is sensed from Thot, just a different ioop. Answer D Discussion rCoI.ECT: See explanation above. Basis for meeting the KA L The KA is matched because the applicant determine the subcooling margin based on the knowledge of the annunciator alarm that is received. - Basis for Hi Cog ______________________ Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANK MNS Bank Q ICICMO48 Development References Student References Provided Lesson Plan OP-MC-IC-ICM page 33: (Rev 16) Lesson Plan OP-MC-IC-ICM page 33: (Rev 16) OP-MC-IC-ICM Obj: 9,10 EPEO74 EA2.01 - Inadequate Core Cooling Ability to determine or interpret the following as they apply to a Inadequate Core Cooling: (CFR 43.5 /45.13) Subcooling margin 401-9 Comments: Remarks!Status 40 1-9 Comments: It is common knowledge that (-) WRT subcooling means subcooling is lost, AandCareNP. This Q is U because of two NP distractors Resolution / Comments: Chose replacement question. See attached file for proposed replacement. Tuesday, July 13, 2010 Page 185 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 64 2564 Tuesday, July 13, 2010 Page 186 of 294

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From Lesson Plan OP-MC-IC-ICM page 33: (Rev 16) 2.3 Subcooling Margin Monitor I Objective # This monitor calculates and displays the subcooling margin of the average of the 5 highest T/Cs and two loop wide range Th and provides alarms for approaching and loss of subcooling. The monitor calculates subcooling using the equation: Subcooling = Tsat Tmeasured

This comparison is made for:

          * (Tsat TIC 5 highest) = subcooling based on the average of 5 highest core
                      -

exit TICs)

         * (Tsat Th ) = subcooling based on loop wide range Th
                     -

The monitor determines Tsat by using wide range loop pressure and then adjusts the value for instrument error associated with the pressure and temperature instruments. The results is a saturation curve similar to the Data Book Curve which the operator uses if the instrument is inoperable. There are provisions for increasing the instrument error in the event containment pressure increases above 3 psig (Sp). This provision is not used at present since the pressure transmitters are located outside containment and would not be exposed to a hostile environment. Inputs to the subcooling monitor are: (refer to Drawing 7.2)

wide range loop pressure (loop D for Train A ICCM and loop C for Train B) from the 7300 Process Control System.
average of the 5 highest T/Cs from the core exit thermocouple monitor calculation
wide range Th (Train A ICCM uses loops C and D while Train B uses loops A and B ) from the 7300 Process Control System
phase B contact is used to indicate that containment pressure is greater than 3 psi. This provides the monitor with the capability of changing the Tsat curve to account for post accident instrument uncertainty however, this function is not used since the displayed curves are the post-accident values.

The subcooling monitor provides the following Control Room Annunciators:

1) Subcoolinq Margin Alert (AD2-D4) This alarm is driven by the train A ICCM only. It will alarm under the following conditions:

2°F subcooling from average of 5 T/Cs or either WR Th

2) Loss of Subcooling (AD2-D5) This alarm is driven by the train A ICCM only.

It will alarm under the following condition:

0°F subcooling from the average of the 5 highest T/Cs or, 0°F subcooling from either WR Th

From Lesson Plan OP-MC-IC-ICM page 33: (Rev 16) Objective # 10 The subcooling monitor has a range of -35°F to +200°F. A positive subcooling margin indicates the coolant is below the saturation temperature (subcooling of the coolant by the number of °F indicated). A negative subcooling margin indicates the coolant is above the saturation temperature (super heated ). The numerical value indicates the degrees of super heat. However, since the range only goes to -35°F, steam superheated greater than 35°F will still only read -35°F. This could lead the operator to erroneously assume that the transient has stabilized. In this situation, the operator should refer to the core exit TIC temperature to see if superheat is still increasing. 0°F subcooling indicates the coolant is at saturation (possible two phase mixture). 2.4 Control Room Display 2.4.1 Controls and General Description Objective # 11 I The ICCM information is displayed on two independent displays (one per train) located on MC-2 (refer to Drawing 7.3). The displays are microprocessor based receiving their data from the ICCM cabinets located in the cable spreading room. The processed data is displayed on the plasma display screens and is updated every 2 seconds. In the event the processor fails, the data will be stored in non-volatile memory and the display freezes as it was at the time of failure. If the failure lasts longer than 10 seconds, a Data Link Failure page will be displayed (refer to Drawing 7.24) and the associated train annunciator ICC MONITOR TRN A(B) TROUBLE on 1(2)AD2-E6 will be in alarm.

There are three components for each display system.

1. the electronics package mounted on the floor behind the control boards

2. the plasma display

3. the key pad

OP/IIA:6l00/OlO C Annunciator Response For Panel 1AD-2 Page 31 of 56 Noxnenclatiue: Window: SUECOOLING MARGIN ALERT Serpohit It NOTE: Subcooling inputs are from Loops C and D Hot Leg RTDs or In core thermocouples. Origin.: Jnadcquate Core Cooling Monitoring Cabinet A Probable Cause:

Degraded NC System pressure or excessive NC System T for the NC System pressure that exists Invalid output from ICCM cabinets Automatic Action.: None Immediate Action.: IT alarm valid and subcooling decreasing. go to applicable Emergency Procedure.

Supplementary Action.: 1. ff alarm invalid, fill ant ICCS Control Room Alarm Data Sheet (located in the bottom drawer aS the NCO Turnover Checklist File Cabinet) prior to resetting ICCM.

2. Reset ICCM from local panel in the cable spreading room (key

98 in Work Control key locker required for cabinel access)

(System will be douai for 1 minute). A. IF alarm does NOT clear, perform the following:

Evaluate system operability
Refer to Tech Specs
Write a Work Request B. Route Data Sheet to the System Engineer.

3. IF alarm is valid due to NC System temperature and or pressure or 1

abnormal ihel temperature, refer to RP.0A:570O0O (Classification of Emergency) for classification of event.

References:

Tech Specs MCM 13 99.73-0016 (ICC S Tech Manual)

End of Response Unit 1

oP!1Iaeloo/clloc Annunciator Response For land LW-2 Page 32 of 56 LOSS OF SUBCOOLING Window: D5 Setpoiut: 07 SiJbcooling OriØ: Inadequate Core Cooling Monitoring Cabinet A NOfl: Subcooling inputs are from Loops C and D Hot Leg RIDs or Incore thermocouples. Probable Cante: Degraded NC System pressure or excessive NC System T n for the NC System pressure that exists

Invalid output front 1CCM cabinets Automatic Action: None Immediate Action: IT alarm valid and subcooling lost, go to applicable Emergency

                        ?rocedure.

Supplementary Action: 1. if alarm invalid, fill out ICCS Control Room Alarm Data Sheet (located in. the bottom drawer of the NCO Turnover Checklist File Cabinet) prior to resetting ICCM.

2. Reset ICCY from local panel in the cable spreading room (key # PS in Work Control key locker required for cabinet access)

(system will be down for 1 minute). A. if alarm does NOT clear, perform the following:

Evaluate system operability
Refer to Tech Specs
Write a Work Request B. Route Data Sheet to the System Engineer

References:

Tech Specs MCM l3P.73 -0016 (ICCS Tech Manual)

End of Response Unit 1

Parent Question: ICICMO48 1 Pt Which ONE of the following will generate a LOSS OF SUBCOOLING (AD2-D5) annunciator in the Control Room? A. 2 °F subcooling on Loop A TH. B. 0 °F subcooling on Loop B TH C. 2 °F subcooling on Loop C TH D. 0 °F subcooling on Loop D TH. Answer 2670 D

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 65 2565 WEO9 EK3.l Natural Circulation Operations

Knowledge of the reasons for the following responses as they apply to the (Natural Circulation Operations) (CFR: 41.5/41.10, 45.6,45.13) Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics. Given the following conditions on Unit 1:

A cooldown is being performed in accordance with ES-O.2 (Natural Circulation Cooldown)
The crew has reached the step in ES-O.2 to initiate a depressurization of the NC system
The crew observes that 2 CRDM fans are running

1. Based on the conditions above, the depressurization continue.

2. The basis, per ES-O.2 Background Document for checking the number of CRDM fans running is to Which ONE (1) of the following completes the statements above?

A. 1. can

2. enhance natural circulation flow B. 1. can

2. prevent voiding in the reactor vessel head C. 1. can NOT

2. enhance natural circulation flow D. 1. can NOT

2. prevent voiding in the reactor vessel head Tuesday, July 13, 2010 Page 187 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 65 256 General Discussion If less than 4 CRDM fans are running at Step 20 of ES-0.1 a more restrictive subcooling margin is required (>100°F instead of >50°F). This is to 1 prevent voiding_in the reactor vessel head. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is correct. Part 2 is plausible because cooling the reactor vessel head provides a small benefit to enhancing natural circulation (i.e. additional head sink along with the SOs). However, the reason for having all CRDM fans in service is to maintain the temperature in the reactor vessel head area below saturation temperature during the depressurization so that a void does not form in the head. Answer B Discussion CORRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible since the step checks to see that all 4 CRDM fans are running. It is reasonable for the applicant to conclude that if less than 4 fans are running the depressurization can NOT continue. The cooldown can continue but the subcooling margin is just adjusted to be more restrictive. Part 2 is plausible because cooling the reactor vessel head provides a small benefit to enhancing natural circulation (i.e. additional head sink along with the SGs). However, the reason for having all CRDM fans in service is to maintain the temperature in the reactor vessel head area below saturation temperature during the depressurization so that a void does not form in the head. Answer D Discussion NCORRECT: See explanatiion above. PLAUSIBLE: Part 1 is plausible since the step checks to see that all 4 CRDM fans are running. It is reasonable for the applicant to conclude that if less than 4 fans are running the depressurization can NOT continue. The cooldown can continue but the subcooling margin is just adjusted to be more restrictive. Part 2 is correct. Basis for meeting the KA The KA is matched because the cooldown and depressurization in ES-0.2 is a transient evolution which could result in loss of natural circulati flow if not managed according to the procedure. The number of cooling fans and subcooling margin is a limitation during the cooldown and depressurization of which the Operator must be knowledgeable. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Learning Objective:

1) EPE0006

References:

( ES-0.2 Background Document (OP-MC-EP-E0)

   )ES-0.2 Natural Circulation Cooldown WEO9 EK3.1 Natural Circulation Operations
                     -

Knowledge of the reasons for the following responses as they apply to the (Natural Circulation Operations) (CFR: 41.5/41.10, 45.6, 45.13) Tuesday, July 13, 2010 Page 188 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 65 Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity inges and operating limitations and reasons for these operating characteristics. 401-9 Comments: niarks!Status 40 1-9 Comments: Stem 2: The basis, per procedure XXX, for checking Include the basis procedure to completely rule out enhancing NC flow since the distractor analysis stated that it was a small beneficial reason. Resolution I Comments: Changed question 2 to The basis, per ES-0.2 Background Document for checking per Lead Examiners recommendation. attached file for revised question. Tuesday, July 13, 2010 Page 189 of 294

Question 65

References:

From ES-0.2 Background Document (OP-MC-EP-E0): STEP 19 if J ANY TIME cooldown rate must be raised to greater than 50°F in an hour, THEN GO TO EPIIIAI5000IES-0.3 (Natural Circulation Cooldown with Steam Void in Vessel). (Continuous Action Step) PURPOSE: To make the operator aware that, if a rapid cooldown is required, another procedure exists which allows for void formation and a continued cooldown/depressu rization. BASIS: From this point onward in ES-O.2, the operator has the option of changing procedures if and when he determines a need to cooldown and depressurize more quickly than at the present rate. Procedure ES-O.3, Natural Circulation Cooldown With Steam Void In Vessel, should be used in this case. The major factors which could require a more rapid cooldown and depressurization than ES-O.2 allows are:

1. Limited condensate storage, or

2. No CRDM fans operating.

STEP 20 Initiate NC System depressurization: PURPOSE: To initiate depressurization of the NC system while maintaining required subcooling. BASIS: The pressurizer pressure should periodically be lowered to maintain the NC and pressurizer pressure-temperature relationship in accordance with the Technical Specifications. The depressurization should be accomplished using pressurizer auxiliary spray or pressurizer PORVs, depending upon whether letdown is in service. To prevent possible void formation in the upper head, the minimum NC subcooling based on core exit T/Cs should be maintained. The depressurization limit is repeated prior to the actual depressurization attempt. This will reinforce the limit to the operator performing the evolution.

From ES-O.2: MNS NATURAL CIRCULATION COOLDOWN PAGE NO. EP/1/AI5000IES-Q.2 16 of 35 Rev. 10 UNIT 1 I ACTION/EXPETEL F:ESPONSE I RESPONSE NOT OBTANEP.

20. InitIate NC System depressurization:

a. Check all CRDM fans ON. - a. Perform the following:

1) Maintain NC subcooling based on core exit T/Cs greater than 100cF.

2) GO TO Step 20.c.

b. Maintain NC subcooling based on core exit TICs GREATER THAN 50F.

c. Check letdown - IN SERVICE. c. Perform the following:

1) Depressurize using one Pzr PORV while maintaining required subcooling.

2) GOTOStep2l.

d. Depressurize using NV aux spray while maintaining required subcooling PER EP11JA!5000/G-1 (Generic Enclosures), Enclosure 3 (Establishing NV Aux Spray).

21. ContInue NC System cooldown and depressurization:

a. Maintain cooldown rate based on T-Colds LESS THAN 50 °F IN AN

HOUR.

b. Maintain subcooling requirements of b. Perform the following:

Step 20.

1) Stop depressurization.

2) Restore required subcooling.

c. Maintain NC temperature and pressure within limits of Data Book curve 1 .6.a.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 66 2566 GEN2. 1 2.1.25 GENERIC Conduct of Operations

                    -              -

( onduct of Operations bility to interpret reference materials, such as graphs, curves, tables, etc. (CFR: 41.10/43.5 / 45.12) Given the following conditions on Unit 1:

A Small-Break LOCA has occurred
Attempts to mitigate the event have been unsuccessful
Core Exit Thermocouples are 630°F and STABLE
Subcooling is (-)5°F and STABLE
A and B NC pumps have been secured
C and D NC pumps are running In order to satisfy the requirements for the Critical Safety Function for Core Cooling, which ONE (1) of the following is required Reactor Vessel DIP based on the conditions above?

REFERENCE PROVIDED A. Train A 15%, Train B 15% B. Train A 15%, Train B 23% C. Train A 23%, Train B 15% D. Train A 23%, Train B 23% Tuesday, July 13, 2010 Page 190 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 66 L 2566 eneraI Discussion In the scenario given in this question, the applicant is presented with a set of conditions to evaluate in order to determine if a challenge to CSF for core cooling is being challenged. F-0 for the Core Cooling CSF requires CET, (<1200) given as 630 deg. Status of subcooling, given as(0 and stable). Status of NCPs (C & D in operation), another check of CET (<700). The next check in the evaluation of this CSF is a check of Reactor Vessel DIP. The applicant is provided a reference in order to determine this value. (F-0 Page 5 of 11). For the given pump combination, the required D/P is 15% for Train A and 23% for Train B. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: The value given for Train A is correct and therefore plausible. The value given for Train B is plausible if the applicant misreads the table and uses the value for C pump being_secured. Answer B Discussion CORRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: The value given for Train A is plausible if the applicant misreads the table and uses the value for A NCP being in operation. The value given for Train B is plausible if the applicant misreads the table and uses the value for the C NCP being secured. Answer D Discussion rn6oRRECT: See explanation above. PLAUSIBLE: The value given for Train A is plausible if the applicant misreads the table and uses the value for the A NCP being in operation. The value_given for Train B is correct and therefore plausible. Basis for meeting the KA KIA is matched because in order to correctly answer this question, the applicant must demonstrate the ability to interpret a given table against a nrovided set of conditions. isis for Hi Cog ibis question is Hi Cog because the applicant must evaluate a given set of conditions and through a multipart mental process, determine a [required value against required parameters. - Basis for SRO only Job Level RO Comprehension BANK MNS Bank Q 1114 Development References Student References Provided Lesson Plan OP-MC-EP-F0 Page 51 (Rev 8) EPIIIAI5000IF-0 Page 5 of 11 OP-MC-EP-F0 Obj 4 GEN2.1 2.1.25 GENERIC Conduct of Operations

                   -             -

Conduct of Operations Ability to interpret reference materials, such as graphs, curves, tables, etc. (CFR: 41.10 I 43.5 145.12) 401..9 Comments: Remarks/Status 401-9 Comments: What reference is provided for this Q? It was not included as part of the Q. This Q is E until verified. Resolution I Comments: Page from F-0 was missing from reference we provided. Added Tuesday, July 13, 2010 Page 191 of 294

_______ FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 66 [to reference. See attached reference material. Tuesday, July 13, 2010 Page 192 of 294

Question 66

References:

OP-MC-EP-FO Obj 4 OBJECTIVES E OBJECTIVE L L P P 0 ,. c OORSR R00 1 State the purpose of each of the six CSF Status Trees. X X 2 Explain the priority system associated with the CSF status X X X trees. 3 Explain the Rules of Usage for Critical Safety Function X X X status trees. 4 Explain the bases for all blocks in the six Status Trees. X X X

From Lesson Plan OP-MC-EP-FO Page 35 (Rev 8) 4.4 Final Plant Status The Core Cooling Tree has defined conditions in all four color priorities as follows: RED PATH There are two conditions which represent an Extreme challenge to Core Cooling:

1. Core exit T/Cs are greater than 1200°F. This condition can only occur if most of the inventory has been removed from the core and heat generation is superheating the steam.

2. Core exit T/Cs are less than 1200°F, but still high enough (>700°F) to show that superheated steam is being generated in the core, and no NC pumps are running and water level in the core is low (< 39% LR RVLIS).

These two conditions will eventually lead to a failure of the fuel/clad matrix barrier, thus it is considered an Extreme challenge. ORANGE PATH There are conditions which represent a Severe challenge to the fuel barrier: Core exit T/Cs are less than 1200°F but subcooling is less than or equal to 0°F and; At least one NC Pump is on, but vessel D/P is less than required, OR No NC pumps are on, core exit T/Cs are greater than or equal to 700°F, but RVLIS shows level in the vessel is greater than 39%, OR No NC pumps are on, core exit T/Cs are less than 700°F, vessel level is less than or equal to 39%, YELLOW PATH If a NC Pump is running and reactor vessel D/P is greater than required 50%, or if no NC Pumps are running, core exit T/Cs are less than 700°F, and vessel level is greater than 39%, but subcooling is less than or equal to 0°F, then the condition is considered to be not satisfied.

EP!IIAI5000IF-O Page 5 of 11 F MNS EP/1 /A/50001F-Q UMT 1 CRITICAL SAFETY FUNCTION STATUS TREES Core Cooling Page 1 of 1

REACTOR VESSEL DIP SETPOINTS FOR DEGRADED CORE COOLING Required REACTOR VESSEL 0/? Number of TRN A TRN B NC Pumps With 1A NC Pump With 1C NC Pump On ON OFF ON OFF 4 44% N/A 44% N/A 3 30% 24% 30% 24% 2 23% 15% 23% 15% 1 16% 10% 16% 10%

Parent Question MNS NRC 1114 Last NRC Exam 2005 A small break LOCA has occurred. Attempts to mitigate the event have been unsuccessful. Approximately one hour after the LOCA first occurred, the operators noticed the Subcooling Margin Monitor in alarm. Given the following conditions on the Inadequate Core Cooling Monitor plasma display:

Core Exit Thermocouples 630 degrees
Subcooling is 0 degrees and stable
A and B Reactor Coolant Pumps have been secured
C and D Reactor Coolant Pumps are running Which one of the following is the required reactor vessel DIP?

Reference Provided A. Train A 23%, Train B 23% B. Train A 23%, Train B 15% C. Train A 15%, Train B 23% D. Train A 15%, Train B 15% Answer 35 A FH-KF, section 3.2

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 67 GEN2.1 2.1.26 GENERIC Conduct of Operations

                -               -

onduct of Operations i(nowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen). (CFR: 41.10/45.12) Given the following:

Operators are performing valve lineups on the Unit 1 Secondary
Several valves are approximately 15 feet above the floor level In accordance with the Nuclear Generation Department Safe Work Practices Pocket Manual, the Operators performing the manipulations can work safely using a securely placed extension ladder and Which ONE (1) of the following completes the statement above?

A. a safety belt with the lanyard attached to a nearby 6 diameter pipe B. a full body harness with the lanyard attached to a nearby 4 diameter pipe C. a safety belt with the lanyard attached to a nearby vertical scaffolding member D. a full body harness with the lanyard attached to a nearby horizontal scaffolding member Tuesday, July 13, 2010 Page 193 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 67 General Discussion In accordance with the Safe Work Practices Pocket manual, Employees shall wear full-body harnesses for fall protection when working on nonwalking/working surfaces (i.e., pipes, beams, hangers, etc.) where a free-fall of greater than 4 ft. exists. Anchor points must be substantial and sufficient to hold twice the weight of a falling person (5OOO lbs.). Examples of acceptable anchor points are I-beam, piping greater than 3 in diameter, structural steel! substantial support structures, etc. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausibe because use of safety belts are discussed in the Safe Work Practices Pocket Manual. However, safety belts are never to be used for fall protection. Also, plausible because a pipe with a diameter of greater than 3 is an acceptable anchor point. Answer B Discussion CORRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausibe because use of safety belts are discussed in the Safe Work Practices Pocket Manual. However, safety belts are never to be used for fall protection. Also, plausible because scaffolding can be used as an anchor point. However, vertical scaffold members can only be used when no other acceptable anchor point is available. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because a full body harness is required. The safety manual does allow the use of scaffold members as anchor points. However, only vertical scaffold members may be used and only when no other acceptable anchor point is available. Basis for meeting the KA he KA is matched because the applicant must have knowledge of the industrial safety requirements (specifically Fall Protection) contained in - ie Nuclear Generation Department Safe Work Practices Pocket Manual. -.______ Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Learning Objectives: -____________________

1) N!A

References:

1) Nuclear Generation Department Safe Work Practices Pocket Manual GEN2.l 2.1.26 GENERIC Conduct of Operations

                   -               -

Conduct of Operations Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen). (CFR: 4l.lO!45.12) Comments: Remarks!Status 401-9 Comments: No comment. Resolution / Comments: Tuesday, July 13, 2010 Page 194 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 67 2567L N/A Tuesday, July 13, 2010 Page 195 of 294

Question 67

References:

From Nuclear Generation Department Safe Work Practices Pocket Manual:

Activating the eyewash handle briefly to verify operation.

6. In case of chemical exposure, flush skin and eyes with cool water for at least 15 minutes, DO NOT RUB!

7. Hold your eyes open with your hands while using eyewash to be sure water reaches the eyes.

8. Remove contaminated clothing after the shower has been activated.

9. Get medical assistance immediately following flushing.

Fall Protection & Walking and Working Surfaces (SWP 3.1 FaIl Protection) General Fall Protection Requirements

1. Employees must be trained in the use of fall protection and have received the practical training.

2. When the fall prevention systems are inappropriate and fall hazards cannot be eliminated or prevented, employees shall control falls by using personal fall arrest systems (i.e., body harnesses, shock absorbing lanyards, self-retracting lanyards, lifelines, ladder safety devices, etc.)

3. Visually inspect fall protection equipment before each use.

4. Employees shall wear full-body harnesses for fall protection when working on nonwalking/

working surfaces (i.e., pipes, beams, hangers, etc.) where a free-fall of greater than 4 ft. exists.

5. Employees shall attach the lanyard as high as possible to avoid striking objects below in the event of a fall. Self-retracting lanyards are available as needed for certain jobs.

26 Note: Shock absorbing lanyards can elongate up to 3 % feet in a fall. This elongation distance must be considered when selecting and using a tie-off point.

6. Safety belts with lanyards shall only be used for restriction of movement or positioning.

Do not use body belts for fall arrest.

7. On vertical lifelines, each worker must have a separate lifeline with a breaking strength of at least 5,000 pounds.

8. Anchor points must be substantial and sufficient to hold twice the weight of a falling person (H5000 lbs.).

Examples of acceptable anchor points are I-beam, piping greater than 3 in diameter, structural steel / substantial support structures, etc.
Examples of anchor points that are not acceptable are instrument lines, small electrical conduit (less than 4 inches in

diameter), plastic piping, valve handles, snubbers, cable trays, instrument trays, hot pipes, etc. Horizontal scaffolding members are un-acceptable and use of vertical supports are to be used only if no other accepted anchorage is available.

9. Workers working in scissors lifts shall be protected by guardrails or personal fall arrest equipment if guardrails are not installed.

10. Wear fall arrest equipment when working on top of tanker trucks or rail cars.

Ladders:

1. When ascending or descending, workers shall face the ladder, use at least one hand to grasp the ladder, and not carry anything 27 that could cause loss of balance or a fall.

2. Extension ladders used to access to roofs, floors, platforms, landings; scaffolds, etc.

must extend at least 3 feet above the access point or be secured at the top and provided with a grasping device to assist workers in mounting and dismounting the ladder.

3. Ladders must be securely placed, held, or tied to prevent slipping and falling.

4. Working load on ladder must not exceed load limits of the ladder.

5. Stepladders are to be used only with the legs fully extended and the spreader bar locked in place. Stepladders must not be used as straight ladders.

6. The top step of stepladders must not be used, except for platform ladders that are specifically designed for that purpose.

7. A harness is NOT required when working from a ladder in an office environment.

8. When working greater than 4 feet off the floor from a ladder:

Always face the ladder
Keep your center of gravity between the rails
Use proper fall protection if it can be done safely and an acceptable anchorage point is available.

Fire ProtectionlPrevention (NSWP 4.2) Fire Prevention

1. Ensure that you know how to recognize and report hazardous conditions and fire hazards associated with the materials and processes to which employees are exposed.

2. Practice good housekeeping in all areas to prevent the accumulation of flammable and!

or combustible material.

3. Keep flammable liquids in approve

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 68 GEN2.2 2.2.25 GENERIC Equipment Control

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quipment Control ( ...nowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5/41.7/43.2) In accordance with Tech Spec 2.1.1 (Reactor Core Safety Limits) Bases, the proper functioning of the (1) AND (2) prevent exceeding the Departure from Nucleate Boiling Reactor Core Safety Limit. Which ONE (1) of the following completes the statement above? A. 1. Rod Control System

2. Pressurizer Safety Valves B. 1. Rod Control System

2. Main Steam Safety Valves C. 1. Reactor Protection System

2. Pressurizer Safety Valves D. 1. Reactor Protection System

2. Main Steam Safety Valves Tuesday, July 13, 2010 Page 196 of 294 O

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 68

..eneraI Discussion In accordance with TS 2.1.1 Basis automatic enforcement of these core SLs is provided by the appropriate operation of the RPS and the steam generator safety valves.                                                                                                                       -

Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 and 2 are plausible because the Rod Control System and Pressurizer Safety Valves perform important functions with regards to plant safety and ensuring that fuel integrity is maintained. However, in the assumptions for maintaining the plant within the design safety limits the Rod Control System and Pressurizer Safety_Valves are not considered. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because the Rod Control System performs important functions with regards to plant safety and ensuring that fuel integrity is maintained. However, in the assumptions for maintaining the plant within the design safety limits the Rod Control System is assumed to not function as designed. Part 2 is correct. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 2 is plausible because the Pressurizer Safety Valves perform an important function with regards to plant safety and ensuring that fuel integrity is maintained. However, in the assumptions for maintaining the plant within the design safety limits the Pressurizer Safety Valves are not considered. Part 1 is correct. Answer D Discussion ORRECT: See explanation above. asis for meeting theKA The KA is matched because the applicant must have knowledge of the Tech Spec Safety Limit basis. Basis for Hi Cog Basis for SRO only EZZ Job Level Cognitive Level QuestionType Question Source RO Memory NEW Development References Student References Provided Learning Objective:

1) N/A

References:

1) TS 2,0 Safety Limit Basis GEN2.2 2.2.25 GENERIC Equipment Control

                   -              -

Equipment Control Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. (CFR: 41.5 / 41.7 / 43.2) 401-9 Comments: RemarkslStatus Proposed replacement for 2010 NRC Q-68. Revised question approved. RFA 06/07/10. Tuesday, July 13, 2010 Page 197 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 68 fl 2568 Tuesday, July 13, 2010 Page 198 of 294

Question 68

References:

From Tech Spec 2.1.1 Basis: eactor Core OLs 82,1.1 5 2D AETY LiMiTs LS 82.1.1 ReaclorCore Ls SACKGRC UXD G 1D Ref. 1:1 requires thai ecmed aocetable fue desigr rn ts are rol exceeded ur1rg steady state oeraiioi, rDrmai ODeraliona: trarsients, an anuclpaled operauoral occurwices AOOsj. This is accornp:she by ha.n a depar1ire tram 9ucleate 000Mg DNBi desgn bas:s. which cc esoords to a 95% proabitty ai a 95% contldeice :eve ithe 595 DNS crterlc9: that DNE wt ot occur and b requIriflg that fuei cenleriire te eraiure slays be ow Ihe 1.eitlg temperature. The resirichois of this L preentoerheaUng oTthe uei and sadcng, as wet. as possiole c adding çierrorallon, thai wo.d resuh in the re ease of fsskin prc.iucts to the reaciorcoclari. OerheaHng of tile £tiei is preer1ed by rnaItainhg the transielt peaic .inear heat rale (LH9) oei the Ieeat tlichfue ceiter1Ire reIrlg occurs. Oert.eat.ng ofhe tuel clatlr* is pre.ented by resdicung lue operator. to a thin the nioieate boilrg regime, where the heat transfer cetTicieit is arge and the ciadirg surtace lePcerature issighByaboe the coo arl saturalioi temperature. Fue, celterIire neitng occurs wten me ocai LHR, or cower DeaMg, l a region. of toe fuei is high enough to cause the tuel cerlesine ternperaiure to reach the rnetilg po it of the tuci. Expar.son of me peuel upon cenierine rnetiig may cause the pe etlo stress the cJaddiig 10 the pcir of fa iure, ai owing an uncoitroited release of aclivity to the reacIor cooant. ODeralian aboe me couldary citric nucleate cot ng regime coud resu in excessive cladding ten peraluro cecause of the oisct of 3N8 and me resutant sharp re-duct or in heat Iransfer coefrcenL Ms de the steam rim, hgh claiing leiperalures are reached, and a cadd.flg water izircolium w.iteri react on may taKe iace. The cheni.cai reaction resuits in oxidaUor of the fuei c.adding to a etruclLJ.rairy weaker fomi. This weaker form -ray lose Is integr.ty, resuMing i an urxxiritroiled release of actJity to me reactor cooar.i. The proper icicnir* or the Reaccr Projection Tern iP and steam geieraIcrsaeiy Qaives creeits iclotori of The reactor core Ls. MoGuire Units 1 and 2 521.1-1 Rean No.51

Reactor Core SLs 9 211 OA3EG APPLICABLE The tue claddlrg must not sustalr dama;e as a result oTnormal SAFETY ANALYSES opra1lon ana AOOs. TM reaetor core SLS ar etablsneD to prac uae vlolaton of the to. owirg tust desl; enera:

a. Th&e mus-l be at lea si 95% procabllty ala 95% cenT deice evel the ONE crr.efl] that the Pot rod In th core noes im experleice tN5: an

b. me bat fuel pellet Ifl the core must lot C xpererce cer:ehlre tue niehlrg.

The Reactor Tra SyteT seo.rts i:Ret. 2, n Donblrator with all the LCO5, are Ueslqred te areit any a,tc aate corn ainatian of imserl acriltors for Reactor Coo art OysternJRCSileTaerature, ROt flaw Rate,si, aressurearJTHERMA_ POWER lewd tat would resLi Ir a eaar1Lre fran iticleate coil nq ratio IENBR of less tan the JN5F limit ari areWde tne exbter9e rffaw nfitabllhiee. Autanatt eifarcenet otthese reactor acre tLs Is proi lied by the aparcarlate oaeratan of tis RPS ard the stern gererat-ar saretv atQes. The as represent a desigl reqL rernet for estab isnirg ihe RPS trp setpo1lts Ideriured preiousty. LOC Iti, ROS Pressure, Temperature. aba F ow Depaittre from rludeaie Eaitmg (ONE: umns: orthe as-surTled ir tat cendtions U the safety alalyses as lncated n tile LIPSAR, Ret 2 plovlde yore resIrlctle limits- to ensue that the as are rot exceeded. SAFEfl LIMITS The FIgure pro1ded i the COLR shows the oci of aoirtz of FractIon of Rated Thermal p ower, RCS Pressure, and average temperature for wi oh me fur Lrnuriu DN s not les5 than the safety aiaiy&es limit that ruei centerine temperature rernans belrj.w melIrg tha7 the average enthalpy In the bct leg s less thar Or equa to the eithalpy of satu rated Iquid. and that the exit qualIty s wIthIn the isuns deilned by the DNSR coirela tan. The reactor core &s are e stabllshed to erecJu.de 1oIatla n of me tot awirg fuel deegn criteria: 95%

a. There must be ai least 35% arobabt tb ata coirderce evel (the 95 / 95 DNE crttertai that the bat tue rod ir me core does rot excerlence ONE; and

b. There must be ai least a 95% arotabl ty at a 95%

conildence evel that lIs hot luel peet lithe core does rot expererae certefllre luel rneitng The reactor core C:Ls are used to dellre tre wa flols RPt fulotlons. suan that the abobe ar tera are 5311sf Cd t urirg steady state operauon. narrnai McGulre Uiits 1 ard2 9 2.1.1-2 Re-.sanNo. 51

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 69 - 569 GEN2.2 2.2.42 - GENERIC Equipment Control

quipment Control bility to recognize system parameters that are entiy-level conditions for Technical Specifications. (CFR: 41.7/41.10 / 43.2 / 43.3 /45.3) With Unit 1 operating at 100% RTP, which ONE (1) of the following exceeds the limits of Tech Spec 3.4.13 (RCS Operational Leakage)? A. 6 GPM identified leakage B. 0.5 GPM unidentified leakage C. 140 GPD tube leakage in 1C SG D. 356 GPD total primary-to-secondary leakage through all SGs Tuesday, July 13, 2010 Page 199 of 294

______ ___ ___________ ___________ ____________ FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 69 2569

   .meral Discussion In accordance with TS 3.4.13 the limiti on operational leakage is:

a No pressure boundary LEAKAGE;

b. 1 gpm unidentified LEAKAGE:

c. I Ogpm identified LEAKAGE;

d. 389 gallons per day total primary to secondary LEAKAGE through all steam generators (SGs); and

e. 135 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant confused the unidentified and identified leakage spec because this leakage exceeds the unidentified leakage. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant confuses the unidentified leakage spec with the pressure boundary leakage spec because this exceeds the pressure boundary leakage_limit.

                                                                                                              -_____

Answer C Discussion CORRECT: See explanation above. Answer D Discussion See explanation above. AUSIBLE: This answer is plausible if the applicant confuses the total primary to secondary leakage spec with the leakage through one SG because this exceeds the allowable leakage through 1 SG. - - Basis for meeting the KA The KA is matched because the applicant must know the limits for RCS Operational Leakage which constitute entry conditions for Tech Spec 3.4.13. Basis for Hi Cog z. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memory BANKJ MNS Exam Bank Q# PSNCO66 J Development References Student References Provided REFERENCE T.S. 2.0 safety limits. GEN2.2 2.2.42 GENERIC Equipment Control

                  -              -

Equipment Control Ability to recognize system parameters that are entry-level conditions for Technical Specifications. (CFR: 41.7 / 41.10 / 43.2/43.3 / 45.3) 1-9 Comments: Remarks/Status 401-9 Comments: No comment. Wednesday, July 14, 2010 Page 200 of 294

__ FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 691 Resolution / Comments: Tuesday, July 13, 2010 Page 201 of 294

Question 69

References:

From TS 3.4.13: ROS Operational LEAKAGE 3,4.13 3,4 REACTOR COOLANT SYSTEM (RCS) 3413 RCS Operaffonal LEAKAGE LCD 34.13 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE:

b. I gprn unidentified LEAKAGE;

c. 10 gpn identified LEAKAGE:

d. 389 gallons per day totat primary to secondary LEAKAGE through all steam enerators (S Gs: and

e. 135 qallons per day primary to secondary LEAKAGE through any one steam generator (SG:.

APPLICABILITY: MODES 1, 2. 3, and 4. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCS Operational A. 1 Reduce LEAKAGE to 4 hours LEAKAGE not within .vithin limits. limits for reasons other than pressure boundar-v LEAKAGE or primary to secondary LEAKAGE.

6. Required Action and 6.1 Be in MODE 3. 6 hours associated Conipktio n Time of Condition A not AND met.

6.2 Be in MODE S. 3n3 hours OR Pressure boundary LEAKAGE exists. OR Primary to secondary LEAKAGE not within limits. McGuire Units 1 and 2 3.4.13.1 Amendment Nos. 2371 219

PARENT QUESTION: PSNCO66 1 Pt Which ONE (1) of the following NCS leak rates at normal operating pressure and temperature is within allowable limits for continued operations per the plant Technical Specifications? (Consider each leak rate separately; assume there is NO concurrent leakage; assume unit in Mode 1). A. 2 gpm unidentified leakage. B. I gpm identified leakage from I NV-6 to the PRT. C. 394 gpd total steam generator tube leakage. D. 9 gpm identified seat leakage thru RHR suction isolation valve 1 ND 1 B and 1 ND-2A to the RHR system. Answer 667 B T.S. 3.4.13 and 3.4.14. Distracter D is PIV leakage, maximum per T.S. 3.4.14 is 5 gpm. KA Nos. I Importance 002G2.1.33 I RO 3.4, SRO 4.0

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 70 257O GEN2.3 2.3.14 GENERIC Radiation Control

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adiation Control

...nowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. (CFR: 41.12 /

43.4/45.10) Given the following conditions on Unit 1:

The NV system is being aligned for startup
The procedure being used calls for independent verification of a single valve located in a room with a general dose rate of 130 mREM/hr
Estimated time to independently verify the valves position is 10 minutes
There are no known hot spots in the area
There is no airborne activity in this room
The room has no surface contamination areas
Assume any necessary approvals are obtained In accordance with NSD 700 (Verification Techniques), independent verification of the valve above (1) be waived because (2)

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

2. the general area dose rate is greater than 100 mREM/hr B. 1. may NOT

2. the general area dose rate is less than 1 REM/hr C. 1. may

2. the radiation exposure for a single verification would exceed the allowable lim it D. 1. may NOT

2. the radiation exposure for a single verification is within the allowable limit Tuesday, July 13, 2010 Page 202 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 70 L General Discussion According to NSD-700, Independent and/or Concurrent Verification may be waived if the exposure to an individual of greater than 10 mrem for a single verification would occur or if dose rate in the room is >1 R/hr. This waiver requires supervisory approval and documentation. Answer A Discussion INCORRECT. See explanation above. lBLE:This answer is plausible because per NSD 700, IV may be waived when dose rate in an area is greater than 1 R/hr, not lOOmR/hr. Answer B Discussion INCORRECT. See explanation above. PLAUSIBLE: This answer is plausible because per NSD 700, IV may be waived when dose rate in an area is greater than 1 RJhr. This [ent is a true statement, but does not correctly answer the question because another limit (> 1 Omr for one IV) is met. Answer C Discussion RRECT._The total exposure would be 21.7 mR which exceeds the dose limit of lOmR for a single verification. Answer D Discussion INCORRECT. See explanation above. PLAUSIBLE: This answer is plausible if the applicant does not recall the guideline of 10 mrem for a single verification criteria or miscalculates potential exposure. Basis for meeting the KA This KA is met because the applicant must evaluate a potential exposure hazard and determine which requirement applies to that potential exposure. Basis for Hi Cog This is an analysis question because the applicant is required to calculate the potential exposure and then apply a limit recalled from memory to orrectly answer the question. 3sis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK 2009 MNS RO Retake Q72 (Bank 1671) Development References Student References Provided _- GEN2.3 2.3.14 GENERIC Radiation Control

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Radiation Control Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. (CFR: 41.12 / 43.4 /45.10) 401-9 Comments: Remarks!Status 401-9 Comments: Change can and cannot to may and may not. Can refers to ability, may refers to permission. Resolution / Comments: Changed question per Lead Examiners recommendation. See attached file for copy of revised question. Tuesday, July 13, 2010 Page 203 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 70 Tuesday, JuLy 13, 2010 Page 204 of 294

Question 70

References:

From NSD 700 (Verification Techniques): VERIfl IL4RD COPY AGAINST WEB SITE IMMEDL4TELY PRIOR TO EACH USE Nuclear Policy Manual Volume 2 NSD 700 700.8 EXCEPTIONS Independent andior Conciurent Verification may be waived under any of the following situations with appropriate supervisoiy approval and documentation:

1. If it would result in a significant personnel radiation exposure as defined below:

a. Indn-idual radiation exposure of greater than 10 mrem for a single verifIcation.

b. Access to an area with a dose rate equal to or greater thai 1 renvhour.

Procedures containing several verification steps. each with high exposures but less than the above exposure limits should be considered for being waived if exposure from verification would exceed 100 nwem per week.

2. In situations that present a significant personnel safety risk.

3. If valves perform a safety function which receive an automatic signal to move to their proper safety position. unless these valves are removed from operability in a manner that would prevent automatic actuation.

4 General vent and drain valves which would NOT prevent a safety-related system from performing its safety function.

5. Under emergency conditions.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2009 MNS RO NRC Retake Examina QUESTION 72 2272 GEN2.3 2.3.14 GENERIC Radiation Control

                  -            -

Radiation Control Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. (CFR: 41.12/ 43.4/45.10) After realigning the NV system for startup, a valve located in a high radiation area requires independent verification. Given the following conditions:

General area radiation levels are 130 MREM I hr
Estimated time to independently verify the position is 10 minutes
There are no known hot spots in the area
There is no airborne activity in this room
The room has no surface contamination areas What are the ALARA requirements related to waiving the independent verification of this valve per NSD 700 (Verification Techniques)?

A. Independent verification may be waived for all valves in high radiation areas until after shutdown. B. Independent verification may NOT be waived until General Area radiation levels are reduced to less than 100 MREM I hr. C. Independent verification may be waived because the exposure to the operator exceeds ALARA guidelines. D. Independent verification may NOT be waived because exposure to the operator will be within ALARA guidelines. Wednesday, July 14, 2010 Page 165 of 174

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2009 MNS RO NRC Retake Examina QUESTION 72 General Discussion -- According to NSD-700, Independent and/or Concurrent Verification may be waived if the exposure to an individual of greater than 10 mrem for a single verification would occur. This waiver requires supervisory approval and documentation. This KA is met because the applicant must evaluate a potential exposure hazard and determine which requirement applies to that potential exposure. This is an analysis question because the applicant is required to calculate the potential exposure and then apply a limit recalled from memory to orrecIty answer the question. Answer A Discussion Incorrect. Plausible because IV in a high radiation could potentially exceed the 10 mrein guidance for a single exposure. However, IV cannot be waived simply because the component is in a High Radiation Area. Answer B Discussion Incorrect. Plausible if the applicant believes that independent verification can not be waived in areas with radiation levels above 100 MREM / hr. Answer C Discussion CORRECT.___________________________ Answer D Discussion Incorrect. Plausible if the applicant does not recall the guideline of 10 mrem for a single verification criteria or miscalculates the potential exposure. Basis for meeting the KA Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension BANK MNS Bank Question ADMOMPNO36 Development References Student References Provided NSD 700 Verification Techniques page 9 (previously OMP 8-2 Verification Techniques) and Lesson Plan OP-MC-ADM-OMP objective 22 related to OMP 8-2. OMP 8-2 was deleted in April 2009 and the ADM-OMP Lesson Plan has not yet been revised to make the change to NSD 700. However, the applicants would have been responsible for the requirements of OMP 8-2 and those requirements did not change with NSD 700. GEN2.3 2.3.14 GENERIC Radiation Control

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Radiation Control Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. (CFR: 41 .12 / 43.4 /45.10) 401-9 Comments: RemarkslStatus G2.3.14 Per Chief Examiners recommendation, replaced distractor A. Distracter A is weak. This is too important of an evolution to be evaluated Changed distractor C to state a reason instead of under these and approved by an NLO. conditions. Rearranged answers for psychometric balance (B to Replace A. A, A to B, D to C, C to D) making C the correct answer. Distracter C: State a reason instead of under these conditions RFA 10/29/09 Wednesday, July 14, 2010 Page 166 of 174

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 71 2571 GEN2.3 2.3.5 GENERIC Radiation Control

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.adiation Control Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.11 /41.12/43.4/45.9)

Regarding the use of Electronic Dosimeters (ED):

If a DOSE alarm setpoint is exceeded, the alarm will (1)
If a DOSE RATE alarm setpoint is exceeded, the alarm will (2)

Which ONE (1) of the following completes the statements above? A. 1. not clear until the ED is reset

2. clear when the dose rate drops below the alarm setpoint B. 1. not clear until the ED is reset

2. not clear until the ED is reset C. 1. automatically clear after 10 seconds

2. clear when the dose rate drops below the alarm setpoint D. 1. automatically clear after 10 seconds

2. not clear until the ED is reset Tuesday, July 13, 2010 Page 205 of 294

___________ ___________ _______________- ______ ___________ __ _____ FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 71 2571 General Discussion This information comes from NSD 507 (Radiation Protection). This is not taught during Generic Rad Worker Training. It is covered during Admin Procedure training in Operator License training. Electronic Dosimeter (ED) Alarms ED Dose and Dose Rate Alarms EDs are programmed during log-on to alarm at a predetermined dose and dose rate. The alarm setpoints are

specified by the RWP. The alarm setpoints can be viewed during EDC log-on and they are also located on the RWP. Set points can also be viewed any time after logging on to EDC by pressing and holding the Dose/Dose Rate toggle switch on the ED for 10 seconds. The alarm setpoints and stay time will be displayed and then will automatically return to dose monitoring mode. The dose alarm consists of an audible alarm and a visual alarm. If the dose setpoint is exceeded the dose alarm will sound and a red light will flash on the ED. The audible alarm and the flashing red light will not stop until the ED is reset. The dose rate alarm automatically resets when the dose rate drops below the alarm setpoint. The ED display will indicate the type of alarm. The ED is also programmed to alarm when it is activated for 16 hours or when RWP specific time stay is exceeded.

Answer A Discussion [T._See explanation above. Answer B Discussion above. PLAUSIBLE: Part 1 is correct. irt2 is plausible because_that is how the DOSE alarm works. Answer C Discussion INCORRECT. See explanation above. °LAUSIBLE: Part 1 is plausible because the 10 seconds is associated with using the DOSE/DOSE RATE toggle switch to view the alarm tpoints. Part 2 is correct. - Answer D Discussion INCORRECT. See explanation above. PLAUSIBLE: Part 1 is plausible because the 10 seconds is associated with using the DOSE/DOSE RATE toggle switch to view the alarm setpoints. Part 2 is_plausible_because this is how the dose alarm works. Basis for meeting the I( ._____

                                                                              .
                                                                                                                    -

The KA is matched because the applicant must be familiar with how the ED alarms works to be able to use an Electronic Dosimeter correctly. Basis for Hi Cog Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Memoty NEW Development References Student References Provided Learning Objectives: RAD-RP#38 eferences: 1)NSD 507 Section 507.7.3 Tuesday, July 13, 2010 Page 206 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 71 A GEN2.3 2.3.5 GENERIC Radiat

                    -             -        ion Control diation Control
  .bility to use radiation monitoring systems, such as fixed radiation monito rs and alarms, portable survey instruments, personnel monitoring equipment, etc. (CFR: 41.11 /41.12/43.4/45.9) emarksIStatus 9

E 40 1-9 Comments: I No comment. Resolution / Comments: I _ Tuesday, July 13, 2010 Page 207 of 294

Question 71

References:

From NSD 507: VERIFY HARD COPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USE NSD 507 Nuclear Policy Manual Volume 2 REVISION 14 VERIFY HARD COPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USE 10 to deactivate the ED after each exit but will deactivate ED at end of shift and return it to storage rack before leaving site. D Complete and turn-in Dose Card even if no dose was received during the entry. D Contact RP regarding problems. 507.7.3 EXPOSURE MONITORING WARNING FLAGSIALARMS A. Radiation Monitoring and Control System Flags To ensure individuals do not exceed dose limits, the EDC computer program provides the following flags as individuals approach their established dose limit: Alert Flag Notification that individual reached 80% or greater, but less than 90%, of established administrative

limit. Individual should notifS his/her supervisor. Individual must receive RP supervision approval to enter a High Radiation Area or Locked High Radiation Area. Exclude Flag Notification that individual reached 90% or greater of established administrative limit.

Individual may not enter the RCA/RCZ until he/she receives a dose extension approved by the Radiation Protection Manager (RPM). B. Electronic Dosimeter (ED) Alarms D Dose and Dose Rate Alarms EDs are programmed during log-on to alarm at a predetermined dose anc

lose rate. The alarm setpoints are specified by the RWP. The alarm setpoints can be viewed during EDC log-oi and they are also located on the RWP. Set points can also be viewed any time after logging on to EDC byl ressing and holding the Dose/Dose Rate toggle switch on the ED for 10 seconds. The alarm setpoints and stay

ime will be displayed and then will automatically return to dose monitoring mode. The dose alarm consists o4 in audible alarm and a visual alarm. If the dose setpoint is exceeded the dose alarm will sound and a red ligh vill flash on the ED. The audible alarm and the flashing red light will not stop until the ED is reset. The dos

*ate alarm automatically resets when the dose rate drops below the alarm setpoint. The ED display will indicat

he type of alarm. The ED is also programmed to alarm when it is activated for 16 hours or when RWP specifi stay time is exceeded.

E If regular monitoring of the ED indicates that the dose alarm set-point will be exceeded prior to completing the job, leave the area and contact RP. Do not wait to receive an alarm before exiting the area. D For some high dose-rate jobs, RP may ask you to exit the work area when the ED accumulates 80% of the dose alarm set-point. D If the ED dose alarm sounds, immediately inform co-workers, exit the RCA/RCZ and call RP. Reentry is not permitted until the alarm is cleared by RP. D ED dose-rate alarms may be anticipated by RP due to higher dose rates in the travel path to the work location OR a worker being in close proximity to a radiation source. Anticipated dose rate alarms shall be discussed during RP brief prior to beginning work. Work can continue following a travel path dose rate alarm providing the alarm clears prior to arriving at the work location. For anticipated dose rate alarms due to proximity to a radiation source, work may continue for no more than two dose rate alarms. If a third anticipated dose rate alarm is received, stop work and notify RP immediately. For unanticipated dose rate alarms (any dose rate alarm that is NOT briefed by RP prior to beginning work) immediately stop work and contact RP. E Notify RP prior to entering RCA or RCZ if you have trouble hearing audible ED alarms. Alternate alarm indicators will be provided. D If the ED malfunctions, immediately exit the RCA/RCZ and call/report to RP with problem ED.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 72 257 GEN2.3 2.3.7 GENERIC Radiation Control

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adiation Control bility to comply with radiation work permit requirements during normal orabnormal conditions. (CFR: 41.12/45.10) Given the following conditions on Unit 1:

The unit has experienced several fuel pin failures
You have been directed to tag out the 1 B NI pump
The lB NI pump room general area is 400 mREM/hr
To reach the I B NI pump room you must transit through a 6 REM/hr high radiation area for 2 minutes and return via the same route
Your current accumulated annual dose is 1000 mREM
An RWP has been written for this job which has your Electronic Dosimeter (ED) alarm set for your EXCLUDE exposure limit Based on the conditions above, what is your MAXIMUM allowable stay-time in the 1 B NI pump room for hanging the tagout to prevent your ED from alarming before you exit the RCA?

A. 30 minutes B. 1 hour C. 1.5 hours D. 2 hours Tuesday, July 13, 2010 Page 208 of 294

_____ _____ ___________ ___________ __ ______ __________ __ _ FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 72

   ,aneral Discussion The Exclusion flag exposure limit is 90% of the 2000 mREM admin limit = 1800 mREM.

The transit exposure is 400 mREM (6000 mREMIhr x 4/60 hr) during transit to and from the job. The allowable exposure before reaching the Exclusion flag exposure limit is equal to the limit minus the transit exposure and the total annual exposure to date. (1800 mRem -400 mREM 1000 mREM = 400 mREM)

Therefore, the_allowable stay time in the 2B NI pump room is 1 hour. Answer A Discussion ______ INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant uses the ALERT exposure limit which would be 80% of the annual admin limit. The jimild be 1600 mREM which would make 30 minutes the correct answer. Answer B Discussion CORRECT: See explanation above. Answer C Discussion iNCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant uses the correct Exclude exposure limit but only calculates the transit exposure in one direction. This would be the correct answer. -_______________ Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant uses the admin exposure limit and only calculates the transit exposure in one direction. us would be the correct answer. asis for meeting the KA The KA is matched because the applicant must be able to determine stay-time in order to comply with RWP requirements. Basis for Hi Cog This is a higher cognitive level question because the applicant must calculate the stay-time based on given information. Basis for SRO only Job Level Cognitive Level QuestionType Question Source RO Comprehension MODIFIED MNS Exam Bank Question RADRPNO3 Development References it References Provided Learning Objectives:

1) RAD-RP #22, 29

References:

1) Fleet ALARA manual_(NSD 507)

GEN2.3 2.3.7 GENERIC Radiation Control

                   -            -

Radiation Control Ability to comply with radiation work permit requirements during normal orabnormal conditions. (CFR: 41.12 / 45.10) 401-9 Comments: Remarks/Status 401-9 Comments: No comment. Resolution / Comments: Tuesday, July 13, 2010 Page 209 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 72 I N/A Tuesday, July 13, 2010 Page 210 of 294

Question 72

References:

From NSD-507: VER,Ifl HARD COPV aca.inr WEB lTE IMXIEDIAflLV PRIOR TO L4CH UE MD 07 Nitchar PolicyMatual Voltiuae to deactivate the LI) after each eat tuc wiLl deactavare zU at end of shalt and return it ti stosage rack beltre leaving sit

Comnlete aM rans-an Data Cast evea if nz dose vat received dining the entry Cannot RP regardn prolileoss.

507J,3 £XPOSURB MONITORIMG W4RN1NG FLAiS/ALARM& S. Raddiehiat \uaiLuaiin titd (VauLt. Sn4ra flag, To ensue individitalt do notexceed dose bait:. rae EDC coaautar )rogin airoudes the fdlowing flags as inin,dni: lppn;eh tl,tlrst?1-Jil-,wI rrhta limit A:en Flag- Nos5carion that andavadaai reached 81% or greates. but less than 90%, ci estab]isId adminirt-ati-re lunar Irdmaal slioialdnoniy bather sapentcr. tnd+/-vidtialmzst receive RI sapenanm awa-ovsl to eater a High Radthnn A.-ea or Laded High Radiator Axes. Exclnde Flag Ncriiicacaon thr irdi-ndta reached 50% or gene:- of established adasazaisrraiva limit.

Inthudual may not eater tie KCAtC £ until he-she recer-es a lose exiennon apgroved by the Radiation Piorection Manager Rfl1). B. ElecirIc Doalmeeer ID) Akirm: El) Dote inc Dosa Rate Ala-ms -EDt are piograaamed cliia-ing log-en to alaraaaat a predereamised dose and ar-tanta Tl la,-m atpnnr rn-a spar.,fiwt lytha RWP The aan cat nrnt sin havieweA ehinrg ElY n-nn and rhoy are alan located cn tho RWP. Pet point tar rho ho nnved mv time thea lapping en to EEC by panstinc ard holdmC the DoteDome Rate toggle switch on the ED for 10 second,. The alas-ta setpomts mod stay Lame tall be cLpl veil vial thin wJl sulwaaaLuAly siam lv ut_n tav.ntavu in-jilt. rln ulose lmn aLt. csf an audible alarm and a ;tua alarm. Ifthe dzte seipoint is exceeded the dose alann will sornd aid a red Light will Oath on them. The audible alarm and the flaslinig red light wail nor stop until the El) is reset. The date rare alarm auronaancaIl cetera amen the dose rate thops below the an aetaomt. The D das,lav trill irdisate the ppe of alarm. The ED as also prozraniaaaedto alaan rvhen it as attiratedfor 16 laura arwkenRW? specific stay rune is exceeded. If e1z raonitoiing of the ED indicates that rIta dose alarm set-point will be eacdadpnorto completing the job. .eave the asea and sontact 2W. Do not trait to receive an alarm aetbre crating tte area.

For same lagh &ae-rate jobs. R? rat- ask yost ta teal the work area when the ED accumulates I 0% of the dust vlaa.u -ke.nat.

a If the ED dos ilana :eund:, hnmediaaetv inform co-werhire, exit he RLAJRCZ anal clI RI. Rc entry L5 aol peaaittcd trill thus oman as titrated by R.

ED dose-rate alanas may be anticanated ha RP due to light dote :-atoa in the travel ; atli to the utah locatan a workes being an dote preacamaty to a radittion torte. Anticipated dose tate aaams shalibe ai-r,,--wl rornig RP bn.fpnrn rat uagtniirg weak Uark ear aarntin,e Infinwang a teavial 7alh fart nte alamprondmg the a-arm dmr: piano to awning at tha avo&loeatsor. For asutapated dote tare alarms dna to pe-oximah- to radaranon source. wink tray continue for no oseit than two dose tnt aimno: a thrd

                 .aaaaicajaied alvsx cut aLvin c Lecta. ca
                                                        . top sutak am uul.sty RPsuaacausdcattlv Its 1                                                              tat ide ala--am (an- dote rate alarm that is tT bueledby EP prior to beginning svork) inamediateh stop work and conner EP.
            *    }Jon.t EP nor to mactn0 RCL         OL SIC   it yauhavo  trouble rearing ouiâle ED alarms. Xltranntr a.orua meucotoas wall be prnddt&

lfnaF.T) natfinwtnrs rnurnütuataiy ascii the SIC A R

1 r? Sri1 ealIsapcrtrn TIP wirhpsnhlara PT) REVISION 1-iR]fl IL4RD COPY aGarcSr WEB SITE 1M%IEDLiTELY PRIOR TO E.iC H USE

Question 72 Parent Question: RADRP N03 1 Pt Units 1 and 2 are at 100% power. Given the following events and conditions:

Unit 2 has experienced several fuel pin failures.
The mechanical seal has failed on the 2B NI pump.
The 2B NI pump room general area is 400 mrem/hr.
In order to reach the 2B NI pump room the worker must transit through a 6 Rem/hr high radiation area for 2 minutes and return via the same path.
The worker has an accumulated annual dose of 400 mrem.

What is the maximum allowable time that the worker can participate in the seal repair on the 2B NI pump and not exceed the EXCLUDE exposure limit for external exposure? A. No longer than 2 hours B. No longer than 2.5 hours C. No longer than 3 hours D. No longer than 3.5 hours Answer 319 Answer: B

Distracter Analysis: The candidate should determine that the exclusion flag exposure limit is 90% of 2000 mrem admin limit = 1800 mrem Transient exposure is 400 mrem (6000mrem/hr x 4/6Ohr). (During transit to and from the job). 400 mrem + 400 mrem = 800 mrem 1800 mrem 800 mrem

             -            =  1000 mrem allowable before reaching exclusion flag exposure admin limit 1000 mrem /400 mrem/hr        = 2.5 hours A. Incorrect: The answer is 2.5 hours.

Plausible: based on using alert flag limit (1600) versus exclude flag. B. Correct: C. Incorrect: The answer is 2.5 hours. Plausible: based on calculating a one-way transit dose. D. Incorrect: The answer is 2.5 hours. Plausible: based on using admin limit (2000) and a one-way transit dose.

FOR REVIEW ONLY DO NOT DISTRIBUTE D

2010 MNS SRO NRC Examination QUESTION 73 2573 GEN2.4 2.4.17 GENERIC Emergency Procedures! Plan

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mergency Procedures! Plan Know1edge of EOP terms and definitions. (CFR: 41.10 I 45.13) Given the following conditions on Unit 1:

A Reactor Trip and Safety Injection have occurred
Main Steam and ALL feedwater is isolated to all SGs
TD CA pump is running.
All SG level instruments agree and indicate as follows:

Which ONE (1) of the following describes the condition of the SGs? A. 1A lB IC 1D Faulted Ruptured Intact Intact [ B. 1A lB 1C 1D Faulted R uptu red Ruptured Intact I C IA lB IC ID Ruptured Faulted Ruptured Intact I D. 1A lB 1C 1D Ruptured Faulted Intact Intact Tuesday, July 13, 2010 Page 211 of 294

__________ ___________________ _______ _____________ _______ ______ __________ _____ FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 73 eneral Discussion [After the reactor trip and safety injection main steam and all feedwater has been isolated. The fact that SO 1A is at 100% level indicates that a SOTL or SOTR has occurred in that SO causing level to increase to 100% (ruptured). I B SO at 0% level indicates that the SO is faulted. The faulted SO would not allow level to increase in the SO even though it is being supplied Auxiliary Feedwater. 1C SG level decreasing slowly is due to it providing steam supply to the TD CA pump. DSO level should be stable since itis bottled up and there is no rupture and no fault. Answer A Discussion -- -________ (iNCORRECT: See explanation above. PLAUSIBLE: The lA SO being faulted is plausible if the applicant confuses difference between a ruptured and faulted SO and how to diagnose each. I B being ruptured is plausible if the applicant does not understand the difference between a ruptured and faulted SO and how to identify them. The lC SO being intact is correct. The ID SO being intact is correct. - Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: The 1A SO being faulted is plausible if the applicant confuses difference between a ruptured and faulted SO and how to diagnose each. B SO being ruptured is plausible if the applicant does not understand the difference between a ruptured and faulted SO and how to identify

   .em.

lC SO being ruptured is plausible if the applicant does not understand the difference between a ruptured and faulted SO and how to identify them. The 1D SO being intact is correct. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: The 1A SO being ruptured is correct. lB SO being faulted is correct. IC SO being ruptured is plausible if the applicant does not understand the difference between a ruptured and faulted SO and how to identify

them.

The lD SO being intact is correct. Answer D Discussion CORRECT: See explanation above. Basis for meeting the KA - This K/A is met because the applicant is required to recall the definitions of terms associated with implementation of EOPs (ruptured. faulted. intact)_and understand how to diagnose_plant conditions relative to those terms. Basis for I-li Cog -________ Basis for SRO only Tuesday, July 13,2010 Page 212 of 294

_____ FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 73 2573 Job Level RO Cognitive Level Memory J QuestionTyp NEW e Question Source Development References Student References Provided OMP 4-3 Rev 31 Pg2 of 35 OP-MC-ADM-OMP Obj: 7 L GEN2.4 2.4.17 GENERIC Emergency Procedures / Plan

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Emergency Procedures / Plan Knowledge of EOP terms and definitions. (CFR: 41.10/45.13)

4O1-9 Comments: kstatus 401-9 Comments: A is NP because 100% in IA and 38% and stable in ID in the stem are diametrically apposed because both indicate stable in distractor A. Replace A. Resolution / Comments: Replaced 1A distracter in answer A with Faulted instead of Intact. See attached file for revised question. Wednesday, July 14, 2010 Page 213 of 294

Question 73

References:

N N L L L L L P P0 OBJECTIVE 0 0 R S R R 00 Concerning OMP 4-3, Use of Abnormal and Emergency X X X Procedures:

Describe the responsibility of licensed operators for maintaining knowledge of and implementation of immediate actions.
State managements expectations for manual initiation of Safequards Actions.
State the expected action RDs and SROs are to take if an automatic action, which should have occurred, failed.
Describe the Operations policy on when Non-procedural blocking of Automatic Safety Actuations could be done.
Given a set of plant conditions, determine if an A.T.W.S.

(Anticipated Transient Without Scram) which would require a manual Reactor Trip has occurred or if a failure of the reactor trip breakers or the automatic trip feature of the reactor protection system had occurred which would require a plant shutdown.

State three subsequent actions that can be taken prior to procedure direction (include conditions that allow these actions to be taken).
State when Adverse Containment Setpoints are used.
Describe the Control Room Team Responsibilities During X X the use of EP/APs.
Define the following items: X X Check, ensure, faulted, ruptured, implement, intact, go to, refer to, per, stable, evaluate.
Describe the rules of use of the Two Column Format Procedure.

ADMOMPOO4

From OMP 4-3 Pg 22 of 35 (Rev 31) 7.16 Selected Definitions S oine words used in the emergency proc eclures have unique meanings. These unique meanins should be understood based upon trainin and experience or by the specific use of the word iii the context of the step being performed. Some words with unique meanings are listed below: Check - to determine present status. (no action) Ensure - to take necessary actions to guarantee that the component or reading is as specified. (Local actions in EPs and APs are only required if step specifies to dispatch personnel though). Faulted - refers to a steam generator that has a secondary break. Ruptured - refers to a steam generator that has a primary to secondary leak (SGTR). Implement - begin a required program or series of procedures. Intact - refers to a steam generator that is NOT faulted or ruptured and is available as a heat sink. GO TO - dis continue use of present procedure and stay in the referenced procedure. The referenced procedure is always entered at the first step unless othenvise specified. REFER TO. PER user is directed to a supplemental procedure enclosure for actions hut will remain in the controlling procedure. Stable - Maintained steady. IF a parameter is being controlled within a desired range. or if a slight trend in either direction is occurring. operator judgment may be used to determine if parameter is considered stable. Evaluate - Appraise the situation. Includes taking action based on evaluation.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 74 GEN2.4 2.4.39 GENERIC Emergency Procedures / Plan

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mergency Procedures / Plan nowledge of RO responsibilities in emergency plan implementation. (CFR: 41.10/45.11) Given the following conditions on Unit 1:

A Site Area Emergency has been declared
A Site Assembly is being conducted in accordance with RPIOIN5700IOI 1 (Conducting a Site Assembly, Site Evacuation, or Containment Evacuation)

In accordance with Enclosure 4.3 (OSM Actions for Site Assembly) the announcement for the Site Assembly shall be repeated (1) until notification that the Site Assembly has been completed and the Site Assembly shall be completed within (2) Which ONE (1) of the following completes the statement above? A. 1. every 20 minutes

2. 30 minutes B. 1. every 10 minutes

2. 30 minutes C. 1. every2O minutes

2. 75 minutes D. 1. everylOminutes

2. 75 minutes Tuesday, July 13,2010 Page 214 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 74 2574 eneraI Discussion In accordance with RPIO/A/5700/0l 1, the Site Assembly should be completed within 30 minutes of initiation and the announcement for Site Assembly is repeated every 10 minutes until notification is received that the Site Assembly has been completed. Answer A Discussion INCORRECT. See explanation above. PLAUSIBLE: Part 1 is plausible since Enclosure 4.3 discusses supervisors calling Security with a report regarding the site assembly approximately 20 minutes after initiation of the Site Assembly. [2is correct. Answer B Discussion CORRECT. See explanationa above. Answer C Discussion INCORRECT. See explanation above. PLAUSIBLE: Part 1 is plausible since Enclosure 4.3 discusses supervisors calling Security with a report regarding the site assembly approximately 20 minutes after initiation of the Site Assembly. Part 2 is plausible as this is the time requirement for activating the TSC. Answer D Discussion INCORRECT. See explanation above. PLAUSIBLE: Part 1 is correct. Part 2 is plausible as this is the time requirement for activating the TSC. Basis for meeting the KA tie KA is matched since ROs typically make the announcements for Site Assemblies from the Control Room and therefore need to know the quirements for Site Assemblies and for making announcements. Basis for Hi Cog Basis for SRO only L Cognitive Level QuestionType Question Source RO Memory NEW Development References feences Provided Learning Objective:

1) EP-EMP #9

References:

1) RP/0/A!5700/0l 1 Enclosure 4.3

)Lesson Plan OP-MC-EP-EMP Section 2.9 GEN2.4 2.4.39 GENERIC Emergency Procedures / Plan

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Emergency Procedures / Plan Knowledge of RO responsibilities in emergency plan implementation. (CFR: 41.10/45.11) 401-9 Comments: RemarkslStatus 401-9 Comments: No comment. Resolution / Comments: Tuesday, July 13, 2010 Page 215 of 294

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2010 MNS SRO NRC Examination QUESTION 74 N/A Tuesday, July 13, 2010 Page 216 of 294

Question 74

References:

From Lesson Plan OP-MC-EP-EMP Section 2.9: 2.9 Site Assembly A Site assembly is an occurrence that warrants the accountability of all personnel on site for reasons of personnel safety or for dissemination of information.

Alert, Site Area Emergency or General Emergency has been declared.
Other plant conditions that, in the opinion of the Operations Shift Manager or Emergency Coordinator, warrant an assembly.

Objective # 9 NOTE: All personnel inside the protected area are to be accounted for within thirty (30) minutes of the initiation of a Site assembly and continuously thereafter.

1. Contact security to inform them a site assembly is being initiated.

2. Turn on outside page speakers.

3. The Operations Shift Manager or designee shall:

a. Sound a 10 second blast of the Site Assembly Alarm.

NOTE: For drill purposes, state This is a Drill, This is a Drill.

b. Using Control Room extension 4262 or 4263, dial 710, pause, dial 80.

Following the beep announce over the Site PA System: This is a Site Assembly, This is a Site Assembly. (Give a brief description/reason for the assembly).

4. Repeat step 3.

5. Continue to repeat step 3, at 10 minute intervals until notification that a site assembly has been completed.

6. Turn off outside page speakers following completion of site assembly.

3.0

SUMMARY

3.1 Review Major Topics Emergency Classification Emergency Response Organization/Facilities Emergency Operations Facility Offsite Agencies Public Alerting/Notification System Access During Emergencies Drill/Exercise Roles Emergency Radiation Exposure

From RPIOIAI5700IOI I Enclosure 4.3: Enclosure 4.3 pt47s7oool1 OSM Actions For Site A;sembl Page 1 of 4 NOIt: 1. All perscnnel inside the protected area are to be acwunted for within thirty (30)niiautes of the icitiation of Site Assembl anti ccntinuously thereafteruntil released or until instructed to reioca:e or evacuate. I All perscnnel outside the protected area and within the owner controlled area should report to their site ssenaby pairn and their supervmor designee within :hirtv (30) minutes ofthe initiation of Site Assembly and continuously thereafter until released or until instructed to relocate or evacuate. (PIP-MO 2-01347)

1. IT a Site Asseably is required and the TS(: is act activated, the Operailons Shift Manager or designec chall perform the following:

1.1 Contact Security at extension 26E3 or 4900 to inform them that a Site Assembly is being initiate& iiO1L: in the event of a card reader tlure. thvistoaGroup Managers are responsible tor accountug Ru ill pnsumitlulidfl Lhen supnvisiuu and calling in a xcpuxt Lu SecaliI1 appiuxucaLely 20 uinntec after initiation of cite acembly. Actions lo be taken in this case are specified in steps 1.7 and LS. 1.2 Confirn that Security aas activated the plant-wide emergency accountability system (card reader system) and that rhe system is fhncticning. 1.3 Turn on outsidc page speakers. 1.4 Sound a 10-second blast of the Site Assanbly alarm. 1.5 Record the site assembly alarm time. Tine 1.6 Record the rime of the Site Assembly alarm from the prev:ous step a: the appropriate space in step hot step jj. to be announced to the site. INTLLS ncm NAME

Enclosure 13 RPIOIA:S700:01 1 OSM Actions ror Site Assembly Page 3 of 4 1.8 For a Drill: dial 710. pause, dial 80. and following the beep. announcec This is a irill. This is a drilL This is a Site Assembly. This is a Site Asseintly. (Gite i brief desrriptinnireisnn fnr ricsernhly/cperl inctilirtinuc) Ailpersonnel are to repit immediately to their asEenlbiypoints. For persons inside the pro:ected area. if you do not know the location of your assembly point, either report ro the Canteen Office Warehouse. or report to the site assembly point in the Adnain Building. For persons outside tie protected area and in the owner controlled area. if you do not know the location of your assembly point, report to the auditorium in building 7422, McCiuire Office Coniplex (MOC). or to the lobby of tuilding 7403. Technical Training Center (TTC). All personnel are to semain at thei: site assembly point until fUrther instructions are given. Assembly start lime is:_______ [PIP M-07-2732. CA. 54) In tie event of a card reader failure, announce: Tie card reader system is not functioning. Division/Group Managers are responsible for accounting for all personnel under their supervision and calling in a report to Security at extension 2688 or 4900 at approximately 20 minutes after assembly start time. 1.9 For an Actual Emergency: repeat steps 1.4 and 1 7 in full, one time. 1.10 [eta Drill: repeat steps 1.4 and 1.8 in full. ane time. 1.11 Contact Security and rejuest that security pcnfbnn a sweep of the discharge canal, tlar natuir trail, and tie beach to evacuate visitors from the owner controlled area. 1.12 For an Actual Emergency: continue to repeat steps 1.4 md 1.7 at 10-minute intervals until notificatioa that the Site Assembly has been completed. 1.13 For a Drill: continue to repeat steps ...4 and ii at 10-minute inerva1s until notification that thc site ascmb1y has been completed. 1.14 Turn off oatside page speakers following completion of site assembly.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 75 ( GEN2.4 2.4.50 GENERIC Emergency Procedures / Plan

mergency Procedures I Plan

bility to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10 / 43.5 I 45.3) Given the following indications on Unit 1:

1AD-2 I C8 (PIR OVER POWER STOP ALERT) is in LIT
Tavg is 578°F and stable

1. Which ONE (1) of the following lists the MINIMUM conditions that will cause the alarm above?

2. Which ONE (1) of the following is the required action for the above condition per OP/1/A/6100/010 C, Annunciator Response for Panel 1AD-2?

A. 1. One PR channel greater than 109%

2. Initiate RCS boration to reduce power B. 1. OnePRchannelgreaterthanlo3%

2. Reduce turbine load to reduce power C. 1. One PR channel greater than 103%

2. Initiate RCS boration to reduce power D. 1. OnePRchannelgreaterthanlo9%

2. Reduce turbine load to reduce power Tuesday, July 13, 2010 Page 217 of 294

_______ FOR REVIEW ONLY DO NOT DISTRIBUTE B

2010 MNS SRO NRC Examination QUESTION 75 2575 eneral Discussion With the indications given, the applicant is presented with a P/R over power rod stop. (C2) The logic and setpoint for this stop is 1/4 P/R> 103%. With the indication given N44 is the only power range> 103%. The correct action per the ARP for 1AD2 C8 is to reduce turbine load to reduce reactor_power. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part I is plausible because there is a PR Hi Flux Hi Setpoint Alert annunciator that will alarm if 1/4 PR instruments is greater than 109% power. Part 2 is plausible because the addition of boric acid to the RCS will add negative reactivity and initially lower reactor power. Answer B Discussion CORRECT: See explanation above. Answer C Discussion I14CORRECT: See explanation above. PLAUSIBLE: Part 1 is correct. Part 2 is plausible because the addition or boric acid to the RCS will add_negative reactivity and initially lower reactor power. Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: Part I is plausible because there is a PR Hi Flux Hi Setpoint Alert annunciator that will alarm if 1/4 PR instruments is greater than 109% power. Part 2 is correct. asis for meeting the KA

   .A is matched because the question tests the applicants ability veri1 the validity of a given annunciator and identifv the correct controls to operate_in   accordance  with the associated ARP.

Basis for Hi Cog This is a higher cognitive level question because the applicant must perform a level of analysis concerning the given indications and determine the cause and select a course of action. Basis for SRO only b_LeveiT RO itie Level Comprehension f BANK Question Source 2009 MNS RO Exam Quesiton #38 L______ Development References Student References Provided Lesson Plan OP-MC-IC-ENB page 51 (Rev 27) ARP for 1AD-2 C8 OP-MC-IC-ENB Obj: 12 GEN2.4 2.4.50 GENERIC Emergency Procedures / Plan

                    -              -

Emergency Procedures / Plan Ability to verif3i system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 41.10 / 43.5 /45.3) 401-9 Comments: RemarkslStatus 0l-9 Comments: No comment. Resolution / Comments: Tuesday, July 13, 2010 Page 218 of 294

____ FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 75 N/A Tuesday, July 13, 2010 Page 219 of 294

Question 75

References:

OP-MC-IC-ENB Obj: 12 12 List the Protection and Control Interlocks (Ps and Cs) 1 X X X X associated with the Nuclear Instrumentation System. (lnclude etpoints and logic) 13 State the purpose of the Wide Range Neutron Detection X X X System. 14 Concerning the Wide Range Neutron Detection System:

    . Describe the operation.
    . Describe the indications and controls.

15 State the purpose of the Gamma-Metrics Shutdown Monitor X X X System. 16 Concerning the Gamma-Metrics Shutdown Monitor System:

   . Describe the operation.
   . Describe the alarms, indications and controls.

x x 17 Determine the validity of indicated reactor power using X X X X alternate indications of power level. 18 Describe the Source Range instrumentation response for X X X X voiding in the core and downcomer region. 19 Concerning the Technical Specifications related to the Nuclear Instrumentation System;

   . Given the LCD title, state the LCO (including any COLR       X X X values) and applicability.
   . For any LCDs that have action required within one hour,     X X X state the action.
   . Given a set of parameter values or system conditions,        X X X determine if any Tech Spec LCDs is(are) not met and any action(s) required within one hour.
   . Given a set of plant parameters or system conditions and     X X X the appropriate Tech Specs, determine required action(s).
   . Discuss the basis for a given Tech Spec LCD or Safety          X

Limit.

SRO Only

From Lesson Plan OP-MC-IC-ENB page 51 (Rev 27) 3.1.4 Setpoints Objective # 11, 12

SR High Flux at Shutdown 1/2 channels 0.5 decade above shutdown counts in

Mode 6 and <5 times shutdown background counts in Modes 3,4&5 (TS Basis 3.3.1).

SR High Flux Level Rx Trip 1/2 channels greater than iü cps.

IR P-6 1/2 channels greater than 10b0 Amps, resets at 7X10

11 amps decreasing. Requires 2)2 channels < setpoint to reset.

lR High Flux Rod Stop C-I 1/2 channels current equivalent to greater than 20%

power.

IR High Flux Level Rx Trip 1/2 channels current equivalent to 25% power.

Power Range Permissive P-b - 2/4 channels 10% power, resets when 3/4 channels < 10% power (Actual values are 10.5% increasing and 9.5% decreasing power).
PR Rx Trip Low Range 2/4 channels > 25% power.

PR Permissive P-8 2/4 channels

                        -                    48% power, resets when 3/4 channels <48%

power.

PR Permissive C-2 1/4 channels> 103% power.

PR Overpower Trip High Range 2/4 channels> 109% power.

PR Positive Rate Trip 2/4 channels > +5% power in 2 seconds.

PR Channel Deviation Deviation between Channels.

PR Upper Section Deviation Deviation between Upper Detectors and the average

of all the Upper Detectors.

PR Lower Section Deviation Deviation between Lower Detectors and the average

of all the Lower Detectors. NOTE: For a complete listing of the Protection Permissive Interlocks and Control Permissive Interlocks (Ps and Cs) see the Reactor Protection Lesson Plan (lC-IPE).

From the ARP for IAD-2 C8 (OPIIIAI6IOOIOIO C) OP/l/A/6100/010 C Annunciator Response For Panel IAD-2 Page 25 of 56 Nomenclature: P/R OVER POWER ROD Window: STOP C8 Setpoint: 1/4 Power Range nuclear instruments at 103% Reactor Power Origin: Bistable in 1/4 P/P. drawers. Protection Sets I. II III. and IV Probable Cause:

Power Range channel in test
Oveipower condition
Instrument malthncticn Automatic Action: Control rods will NOT withdraw in automatic or manual.

Immediate Action: 1. jj alarm is due to test reduce output of channel below Setpoint and place tRod Stop Bypass switch to the channel in test.

2. IF oveipower condition exists, reduce power below 100% Reactor Power.

3. a due to instnunent inalitinction. go to AP/1IA/55 00/016 (Malfunction of Nuclear Instnunentation).

Supplementary Action: IL desired to have Engineering evaluation as to cause for alarm. freeze the Transient Ioiiltor.

References:

UPSAR. Fizure 7-1 (5 of 16)
Drawing MCM-1399-04.27
NSMMG-12126 Eud of Response

Copy of parent question 2009 MNS RO Exam Q 38: 1 Pt Given the following sequence of events:

A Large Break LOCA occurs on Unit 1
All ECCS systems are injecting from the FWST
Safety Injection is reset
FWST level is currently 200 inches An Operator depresses the SS-RESET pushbuttons on the CNTRL PERMISSIVE FOR RECIRC MODE 1NI-185A1 184B switches.

Concerning the following valves:

1 NI-I 85A (RB Sump to Train A ND & NS)
1NI-184B (RB Sump to Train B ND & NS)
1ND-19A (A ND Pump Suction from FWST or NC)
1ND-4B (B ND Pump Suction from FWST or NC)

Which ONE (1) of the following describes what the Operator observes with regards to the automatic operation of the ECCS valves listed above after the SS-RESET pushbuttons are depressed? A. Immediately after depressing the SS-RESET pushbuttons, 1 NI-185A1184B, OPEN AND 1 ND-19A14B CLOSE. B. Immediately after depressing the SS-RESET pushbuttons, 1NI-185A1184B OPEN AND 1ND-19A14B REMAIN OPEN. C. When 2/3 FWST Lo Level Bistables are received, 1NI-185A/184B OPEN AND 1ND-19A14B CLOSE. D. When 2/3 FWST Lo Level Bistables are received, INI-185A/184B REMAIN CLOSED AND 1ND-19A/4B REMAIN OPEN.

2010 MNS SRO Question Worksheets

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 76 [761 SYS003 2.1.20 Reactor Coolant Pump System (RCPS)

( (S003 GENERIC

 .bi1ity to interpret and execute procedure steps. (CFR: 41.10 / 43.5/45.12)

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 (Malfunction 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 1A NC pump #2 & #3 seals

2. only while AP-08 is in effect B. 1. damage to the IA 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 Tuesday, July 13, 2010 Page 220 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 76 2576 _eneral 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 (INV 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 INV-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 EPsto arrive at the correct answer.

Answer B Discussion CORRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: Part I is plausible because this is the basis for closing INV-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 riswer D Discussion CORRECT: See explanation above. PLAUSIBLE: Part I 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. 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 [ute 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): I) 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.

Job Level Cognitive Level I QuestionType J_____________ Question Source SRO Memory NEW Tuesday, July 13, 2010 Page 221 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE -

    -      2010 MNS SRO NRC Examination                                       QUESTION 76                         6 L

,,....ieveIopment References Student References Provided

        ;son Objective:

References:

1) AP-08, Malfunction of NC Pump

2) AP-08 Background Document SYS003 2.1.20 Reactor Coolant Pump System (RCPS)

SYSOO3 GENERIC Ability to interpret and execute procedure steps. (CFR: 41.10/43.5/45.12) Comments: RemarkslStatus 401-9 Comments: No comment. Resolution / Comments: N/A Tuesday, July 13, 2010 Page 222 of 294

Question 76

References:

From AP-08: MNS MALFUNCTION CF NC uri PAGE NO. AP/iA55D08 of 24 C,e I UNIT 1 NC Pump Seal or Pump Lower 3earing Nlalknc:ion Pev. 12 ADT:cwEPE:ED ESPC}TEL P4EEtNSE NOT DBtAflEt

6. Check if scat cooling avaiIable to Perform the folowing:

affected pump: Cl OS me of 1h4 fnlfrxhçy Seal irjecticn established korrnal o 5SF supplvi

1NV-94AC (NC Rinips Seal Ret Cont Inside sell OR OR
XC Lu hernial unniei Liblislied.

                                                                      . 1N\9GD (tC Pumps Seal Ret Cont Outside lsol.

I _b. IELT4NYJIMFSe3IDOOIirgi; restored. IHEN cbser:e Nole prior b Step 7 and GO ID Step 7. _c. RETIJRNTOSte2. NOTE Lp to 24 ,oursof NC pump operatioi nuvç be required bebre seals seat aid opcrate normally after scal maintenance or startup. _7. Check anvNC pump miiibei Iseal Perform thefoloNing: leakoff GREATER HkN OR EQUAL TO

GGPM. NJ It UPJ1:PJhZUWUUlb (Cheni cal and Vclurne Cunnju Sysiri I Chaging. EncIoure 4 10 (Maintaining NC Rinp Seal Lerkdf) g ve guidarce or a:tionis usec to charge seal leakoif flow.

a. IF leakoffslcwly goirg up, THEN ntact s:aton management tar further guhiarce.

b. Coritinuetomoni:orNC pump seal leakoff flew.

_.e IFLTANYflMFscnI atfffhge up to 6 GPhI, THEN GD TO Step 8.

d. QIQStep9

MNS MALFUNCTION OF NC PUMP PAGE NO. AP1l/A15500/08 6 of 24 Case I UNIT 1 NC Pump Seal or Pump Lower Bearing Malfunction Rev. 12 I DTION/EXPETED RESPONSE RESPONSE NOT OETAINEO

8. Stop affected NC pump as follows:

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

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

INC-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 MAI Fl JN1TlON OF NC P1 IMP PACF NO AP/ I INS SOD/OS Enclosure 2- Page 1 of 2 ¶224 UNIT 1 NC Pump Post Trip Actions For #1 Seal Failure Re. iCIIDN/EXPECrEt PESIDNEE aEECNSE NOT O3TA2ED CAUTION Failure ofNumber2 and 3 seals may occur unless the affected NC pump seal return valve is closed between 3 minutes and 5 minutes after slopping pump. This enclosure must be completed even after transtion 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:

e Seal injection eaIablI3hed Nomial r 5SF SuDplv) . 1NV-34AC (NC Pumps Seal Ret Cant Inside Isol) CR

                                                           . 1NV-ibS (NC Pumps Seal Ret Cant

        . KC to themial barier estalished.                       Outside 1301).

b. Exit this enclosure.

3. Check anj NC pump number 1 sea GO TO Step 5.

leakoffficw GREATER THAN OR

EQUAL TOG GPM.

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

MNS MALFUNCTION OF NC PUMP PAGE NO. AP/i /A/5500/08 24 of 24 Enclosure 2 Page 2 of 2

UNIT 1 NC Pump Post TrIp Actions For #1 Seal Failure Rev. 12 ACTION/EXPECTED RESPONSE RESPONSE OT OBTAINED

5. WHEN affected NC pump has been off 3 minutes THEN unmediately perform the folowi ng:

a. CLOSE affected NC pump seal return valve:

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

      . I NV-66A (C NC Pump Seal Return so!)

1 NV-8213 (D NC Pump Seal Return Isol).

b. OPENijofthefoIlowing valves:

     . OPEN 1KC-394A (A NC Pun,p Therm Bar OtIt).
     . OPEN 1KC-345A (C NC Pump Therm Bar OtIt).

OPEN IKC-364B (B NC Pump Therm Bar Otlt).

     . OPEN IKC-413B (0 NC Pump Therm Bar OtIt).

From AP-08 Background Document: APII and 21A155 001008 (MaIfunetio 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 individual 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 cocflng events. The following is an excerpt from DW-94-Q1 1: 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 far the extended loss of ac power in WCAP 1 0541 Rev. 2, identifies that the high temperature two-phase flow through the seal system and

1 seal leakoff line increases the pressure in the #1 lealcoff 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 & 95 keeps individual flowpath open for #1 seal leakoff (through relief to PRT) but isolates it from the \CT, Note: this AP does not address restoring a loss of seal injection or thermal barrier cooling. Other APs address those problems.

REFERENCES:

DW-94-O1 I page 6 of 7 WCAP-10541. Rev.2 Page7of39 Rev2

APII and 21A15500!008 (Malfunction of NC Pump) CASE I STEP 8: PURPOSE: Provide 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-03-1 992 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 .ihile 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 beloi 5%, tripping the affected NC pump, and transition to E-0. 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. PagelOof3S Rev2

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 77 2577 SYSOO5 A2.02 Residual Heat Removal System (RHRS)

bility to (a) predict the impacts of the following malfunctions or operations on the RHRS. and (b) based on those predictions. use procedures o correct, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 /43.5 /45.3 /45.13) Pressure 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. n B. Secure two PORV5 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. Tuesday, July 13, 2010 Page 223 of 294

_______ _________ ______ ______ ________ _______ _________ _______ ________ FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 2577 General 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 satisfy 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 _J

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 spreventing_a PTS_type event. - - Answer B Discussion CORRECT: See explanation above Answer C Discussion INCORRECT: See explanation above. LAUSIBLE: Part 1 is plausible because it is partially correct in that the actions stated would meet LTOP requirements but an additional

  )erable 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): It can NOT be answered solely by knowing < 1 hour Tech Specs. . It can NOT be answered solely by knowing the LCOJTRM 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 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.

Tuesday, July 13, 2010 Page 224 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE B

2010 MNS SRO NRC Examination QUESTION 77 2577 fobLeveif CognftiveLev& QuestionType Question Source Comprehension NEW Development References Student References Provided Learning Objectives:

1) PS-NC #24

References:

Tech Sped 3.4.12 Basis SYSOO5 A2.02 Residual Heat Removal System (RHRS)

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: 41.5/43.5 / 45.3/45.13) Pressure transient protection during cold shutdown 401-9 Comments: RemarkslStatus 401-9 Comments: No comment. Note: The justification on page B3.4. 12-8 of the reference should be before distractors C and D NOT A and B. Resolution / Comments: Revised the justification on reference page B 3,4.12-8 to say This provides plausibility for distracters C and D. Tuesday, July 13, 2010 Page 225 of 294

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

x x x

       . For any LCDs that have action required within one hour, state the action.

x *

       . 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 BASES APPLICABILITY (continued) the pressurizer safety valves that provide overpressure protection during MODES 1. 2. and 3, and MODE 4 above 30D°F. Low temperature overpressure prevention is most critical during shutdown when the P05 is water solid, and a mass or heat input transient can cause a sery rapid increase in RCS pressure when little or no time allows operator action to mitigate the event. The Applicability is modified by a Note stating that accumulator isolation is only required then the accumulator pressure is more than or at the maximum RCS pressure for the existing temperature, as allowed by the RT limit curves. This note pemits 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.O.4.b to an noperabIe LTOP system. There is an increased risk associated with entering MODE 4 from MODE 5 with LTOP inoperable and the provisions of LCD 3.D.4.b. which allow entry 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. Al. Aul,A22.1. A22.2. &3, At Af.1. andA.5.2 6r.k centrifuqal charqinq pumps, or a combination of each. capable of injecting into the RCS. ROS overp ress urization is possible. To immediately initiate action to restore restricted coolant input capability to the RCS reflects the urgency of removing the RCS froni this condition. Two pumps may be capable of injecting into Ihe RCS provided the RHR suction relief valve is OPERABLE with: Coxrect arnwer. one option to allow the 1. RCS cold leg temperature > 174 F (Unit 1), or tetwou1d be 2. RCS cold leg temperature > 89°F Unit 2), or 2 3. RCS cold leg temperature > 74°F and cooldown rate < 20°Fihr (Unit 1), P0KV; or

4. RCS cold leg temperature > 74°F and cooldown rate 50°F/hr (Unit 2).

<:

or

5. r-CRVS secure.h vth associated olock es open oil o.ver rema.e&j&

5, a RCS vent of 4.5 square inches, or McGuire Units 1 and 2 6 3.4.12-7 Revision No. 102

LTOP System 614.12 BASES ACTIONS (continued) 1 RCS vent of>2.76 square inches and two C:PERAELE 8C8 vent shall not be cne of tne two OPERABLE F.Rts:I Thn provides For cases where no reactor coolant pumps are in operation, P05 cold le plausibilit for temperature limits are to be met by monitoring of BOTH the WR Cold Leg dishacten c: and 0. temperatures and ResidLial Heat Aernol he?! FYC2nqer schrcre eratwe limit the flow relief capac iw. Fort ne RIHR relief .ale to be OPER me RHR suction soatioi valves must be open r he lief setpoint at 4sig corsstent ::ith the safetv n.. . The RI-JR suction reef va qes are spnirg loaded, bellows type water relief valves with 1 pressure tolerances and accumulation limits established by Section III of the American Society of £viechanical Engineers (ASMEi Code (Ref. 3) for Class 2 relief valves.. Required Action A. 1 is niodified by a Note that permits tvo centrifugal charging pumps capable of RCS injection for 15 minutes to allow for pump swaps. 6.i,C.1.aredC.2 An unisolated accumulator requires isolation within 1 hour. This is only required Mien the accumulator pressure is at or more than the maxi mum RCS pressure for the existing temperature allowed by the PiT limit curves. If isolation is needed and cannot be accomplished in 1 hour, Required Action C. I and Required Action C.2 provide two options. either of Much must be performed in the next 12 hours. By increasing the RCS teniperatureto> 300°F, an accumulator pressure of 639 psig cannot exceed the LTOP limits if the accumulators are fully injected. Depressurizing the accumulators below the LTOP limit also gives this protection. The Completion Times are based on operating experience that these activities can be accomplished in these time periods arid on engineering evaluations indicating that ari event requiring LTOP is not likely in the allowed times. 0.1 In TiODE 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-s evision No. 102

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 78 2578 A2.06 Auxiliary I Emergency Feedwater (AFW) System

SYSO61 bi1ity to (a) predict the impacts of the following malfunctions or operations on the AFW; and (b) based on those predictions. use procedures to

  .orrect, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 / 43.5 I 45.3 I 45.13)

Back 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 ChkVlv 1CA-37) alarms on the OAC
1CA-37 (#1 TD CAto SIG D)

Based on the above conditions:

1. In accordance with 0P111A16250/002 (Auxiliary Feedwater System), 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 1CA-36 AB (Ui TD CA Pump Disch to 1D SIC Control) and monitor temperature for 15 mm.

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

B. 1. Close 1CA-36AB (Ui TD CA Pump Disch tolD SIG Control) and monitor temperature for 15 mm.

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

C. 1. Close 1CA-38B(U1 TD CA Pump Dischto 1D SIC Isol) and starttheTD CA pump aligned for recirculation to the UST.

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

D. 1. Close 1CA-38B (Ui TD CA Pump Disch to 1D 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.

Tuesday, July 13, 2010 Page 226 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 78 2578j General Discussion The consequence of the situation described would be overheating of the ID 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 OP/11A162501002. 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 ICA-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, closing this valve renders inoperable. the pump Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 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 RRECT:_See explanation above. Answer C Discussion INCORRECT: See explanation above. 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. art 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 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 Part 2 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 1 of this question matches the part b of the KA regarding using procedures to correct, control, or mitigate the consequences. The pLocedure in this case is OP/l/Ai6250/002, 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 [pability of the associated AFW pump. _______________________ Basis for SRO only Part I 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)(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 pump piping. Check valve leakage is discussed in the systems lesson plan and the methods to cooldown the TD CA pump are mentioned in ieral terms (i.e. close the discharge valve or start the pump). However, the applicant must have detailed knowledge of the OP to

..iscrimminate which method is used FIRST. Since this is an infrequently performed evolution, the actions in the procedure are directed by the CR 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.

Tuesday, July 13, 2010 Page 227 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE B

2010 MNS SRO NRC Examination QUESTION 78

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

is 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 hour 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.

L JobSROLevel Cognitive Level Comprehension QuestionType NEW

                                                                    -                              Question Source Development References                                                                         Student References Provided TS 3.7.5 OP/1/A!6250/002 Auxiliary Feedwater System SYSO61 A2.06 Auxiliary / Emergency Feedwater (AFW) System
                    -

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 / 45.13) Back leakage of MFW 401-9 Comments: RemarkslStatus 401-9 Comments: Must reference the procedure in the stem to fully meet the 2nd part of the KA. Resolution / Comments: Revised question 1 in the stem to read In accordance with OP/l/A16250/002 (Auxiliary Feedwater System), what method would FIRST be used to reduce the temperature at the check valve? See attached file for revised question. Tuesday, July 13, 2010 Page 228 of 294

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 hours 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 SR3.7.5.1 -NOTE o applicable to automatic valves when THERMAL ER is < 10% RTP. 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 NOTE Not required to be performed for the turbine driven AFW pump until 24 hours 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 OPII 1A162501002, Auxiliary Feedwater System: Enclosure 43 op/i /A6250/002 Reducing Turbine Driven CA Pump Piping Page 1 of 3 Temperature

1. Limits and Precautions None

2. Initial Conditions 2.1 3W System isolated from S/Cs per OP/1/A16100,SO-5A (B. C. D) (Draining S/Cl

l. lB. IC. 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 inopemble. SRO 3.3 Close control valve on affected lines:

1CA-MAB (Ui ID CA Pump Disch to lA S/Cl Control) cv
1CA-52AB (UI ID CA Pump Discli to lb St Control) cv
1CA-48AB (01 ID CA Pump Disch to lC S/Cl Control) cv
1CA-3AB (121 ID CA Pump Disch to 1D S/Cl Control) cv Q 3.4 Monitor temperature for 15 30 minutes.

3,5 IF temperatures remain high after 15 -30 minutes, close isolation valve on affected lines:

1CA-66AC (121 ID CA Pump Disch to l S/Cl J.sol) cv
1CA-S4AC (Ui TD CA Pump Disch to lB S/Cl Isol) cv
ICA-50B (01 ID CA Pump Disch to lC S/Cl Isol) cv
1CA-38B (01 ID CA Pump Disch to 1D S/Cl Isol) cv Unit 1

7 Enclosure 4.5 oP/i...vo25o.o2 Reducing Turbine Driven CA Pump Piping Page 2 of 3 Temperature NOTE: When opening valves 1CA-36. 4S. 52. and 64 from the local panel, the controller needs to be opened 4 5 more turns once 100% is reached to minimize the amount that the

valves drill close and back open upon returning controller back to control room (A-Remote). 3.6 AT[ER lemperatures have returned to normal. ensure open:

IC-64AB (Ut TI) CA Pump Disch to 1A S/Cl Control) cv
1CA-52AB (171 ID CA Pump Disch to lB S/Cl Control) cv
ICA-4SAB (171 ID CA Pump Disch to 1C S/Cl Control) cv
ICA-36A3 (Ui IT) CA Pump Disch to 11) S/Cl Control) cv
1CA-66AC (Ui TI) CA Pump Disdh to IA S/Cl isol) cv
1CA-S4AC (171 TD CA Pump Disch to 13 S/Cl Isol) cv
1CA-SOB (171 U) CA Pump Disch to 1C SiClIsol) cv
ICA-38B (Ui ID CA Pump Disch to 1D SCl J.sol) cv 17 Check the following stable:

D M1A1439 (171 CA Temp at Cli iv 1CA-65) C] M1A1421 (171 CA Temp at Cli Vlv ICA-53) C] M1A1294 (171 CA Temp at Cli Vlv 1CA-49) C] M1A1276 (171 CA Temp at Cli 1k ICA-37) Unit 1

Enclosure 4.5 oWiA.25o.o2 Reducing Turbine Driven CA Pump Piping Page 3 of 3 Temperature 31 II increasing temperatures indicates check valve leak by. perform the following: 31.1 Notify System Engineer. Person Notified Date Time 312 Evaluate operating CA Pumps to cool CA System piping. 19 Ensure TUEB 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 1

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 79 Ei SYSO76 A2.O1 Service Water System (SWS)

bi1ity to (a) 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 /45/13) Loss of SWS Unit 1 is operating at 100% RTP with B train components in service with a normal RN system alignment. 1 B RN pump amps are swinging. The following annunciators are in alarm:

B RN Pump Suction Lo Press
B RN Pump Discharge Lo Press Which ONE (1) of the following is the required response to the above conditions based on the implementation of AP-20 (Loss of RN)?

A. Implement Case 1, Loss of Operating RN Train Swap alignment to the Nuclear Service Water Pond and place the 1A RN pump in service B. Implement Case 1, Loss of Operating RN Train Place the 1A RN pump in service and remain on Low Level Intake C. Implement Case 2, Loss of Low level or RC Supply Crossover Swap alignment to the Nuclear Service Water Pond and place the 1A RN pump in service D. Implement Case 2, Loss of Low level or RC Supply Crossover Place the 1A RN pump in service and remain on Low Level Intake Tuesday, July 13, 2010 Page 229 of 294

________ ________________ ________ ______-_______ _________ FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 79 2579 General Discussion The swinging amps, lo suction pressure, and 1 discharge pressure indicate that the RN pump is cavitating and that the Low Level Intake (LU) has been lost. Hence Case 2 for Loss of LLI is the appropriate procedure to implement. If the lo suction pressure alarm was not present, it would indicate that the LU had not been lost and entry into Case I Loss of Operating RN Train would be appropriate. lAW Case 2, the Operators will swap to the Standby Nuclear Service Water Pond (SNSWP) and then swap RN pumps. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant does not comprehend that the pump suction lo pressure alarm prompts implementation of ase_2.The other actions are reasonable as they would be the correct actions if Case 2 were implemented. Answer B Discussion Ib0RRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant does not comprehend that the pump suction lo pressure alarm prompts implementation of [çse_2. The other indications alone would prompt entry into Case I The remaining actions are correct for Case I. Answer C Discussion -_______ CORRECT: See explanation above. Answer D Discussion - INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because Case 2 is the correct case to_implement._The actions are correct if Case I had been_inptej Basis for meeting the KA - KA is matched because the candidate must evaluate the provided conditions to determine the impact on the system and determine from those ,ridications which is the appropriate procedural strategy. ç asisforHiCog 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 solely by knowing systems knowledge. The question required the applicant to have detailed knowledge of AP-20 content.

2) The question can NOT be answered by knowing immediate operator actions. There are no immediate actions associated with AP-20.

3) The question can NOT be answered solely by knowing entry conditions for AOP or direct entry conditions for EOPs. The question requires the applicant to have detailed knowledge of AP-20 content.

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 step sequence not sequence of events within the procedure. The question requires the applicant to have detailed knowledge of AP-20 content.

5) The question requires the applicant to analyze plant conditions and determine which section of the procedure to use to mitigate the consequences of the accident. Once the applicant determines which section of the procedure is required, they must recall from memory the actions that are directed to mitgate the accident (detail knowledge of procedure content). Therefore. this is an SRO level question.

Job Level - Cognitive Level QuestionType Question Source SRO Memory BANK MNS 2009 NRC Q80

Development References Student References Provided AP/I/A15500/20, Loss of RN, rev. 25 pages 2 and 29 35. - tS076 A2.01 Service Water System (SWS)

Ability to (a) 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 / 45/13) Loss of SWS Tuesday, July 13, 2010 Page 230 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 79 i1-9 Comments: arksIStatus Proposed replacement for 2010 NRC Q-79. Replacement question approved. RFA 07/06/10 Tuesday, July 13, 2010 Page 231 of 294

Question 79 Proposed Replacement

References:

From AP-20: MNC LOSS OF RN PAGE NC. b PuiAcocJ bf 111 T3NIT 1 I Lcss of Op;raung RN Tralr

                    *K.tJUNf&X ir:c flatntiIit S. Symptoms                                                  fcr   etr      into

A RN PMP DISCHARGE LO PRESS alarm
PMP DISCHARGE LO P alarrn
A RN PUMP ABNORMAL FLO alarm
6 RN PUMP ABNORMAL FLO alarm
RN Plori-Essentlal Header pressure. low
RNpumptrtpped
Indications oteignrncant RN pump cavitation (flow, pressure, anips swinging)
Entry Lnt this AP has boer spaeLtlsd by another procadurs.

C. Operator Action8

1. Check tar pote.nltaI loss or ILl as IF Ion of LLI suction supply to RN follows: pumps Is believed to han occurred, THEN GO TO C ass Ii [Loss of Low CII ec& ur 12 RN pump:s mal are Level or RC supply Crossover].

aligred to LLI OPERATI S

PROPERLY. Check suciior fluszath AVMLAB LE.

2. Anaouncs occtrrenon page.

3. Check If signIficant RN pump cavItation _GOTOStepC.

pressure, amps swingIng] I-S

CCC 1JRRIN G

kNS LOSS OF RN PAGE NO. AP/1S55OC2] Oasa 3] or iii Rev. 29 TNT 1 LosE olLow Level or 90 Supply Crossoeer I UN/Lx1ECrKU :tL I thi Al FILL.

6. Sprptoms ARNPMP SUCTION LOPRESSaIarrn 4 -fl ftN PMP SUCTION LC PfttS alai iii A RN PMP DISCHARGE LO PRESS atarin a 5 RN PMP DISCHARGE LO PRESS aLarm

tA. RN pump amps indicate low
16 RN pump amps indftate low
VLsuaI obweriatlon of Cowans Ford Dam rallure wilt potenhti loss at LLI
NotIfication from plant personnel of potential lows or LU
Indications at signllicant RN jump cavItation (flow. presaure. amps BwIngIng) on both Lnltr
Entry Into this AP tas been specified by another procedure.

MNG LO OFRN PAGE NO. case ii 31 & Ill UMI 1. Lose ol Lc LeeI or RC upPIy Crsser

              ,.:LLuNILX tL LU        *4                                    IL: .:      IJ  . 1ND C. Opartor Actions CAUTION            St&pa 1 through 18 are time crttlcal nd miiaL be peIlorTned wI(hou delay.

NOTE Case IL iLo oLow LeeI or RO Guply Crassoer) Is derllca. 19 noth wilts AP2 an 011y needs to be competed ore. iflarE-i RN ates car be pcrated from either unirs cnlro swItCfl.

1. OPEN ORN-B CTraEn lB & 2e SNSWP GO TO StepL uppiyj.

2. Check LI failure of Cowafta Ford Darn or Perform the toIIowIng:

LLI BELIEVED TO HAVE OCCUR RED

a. IF tiIs AP erlered ltom P.WA57O,DD7 Ea quae, THEN GO TO step 7.

                                                               . FE this AP erlered ltorn APA55E<l,7 e.curlty Eents:., ThEN .V ID step 8.

QJ[OStfip7.

_4. Check LIeIinIflcant RN pLLIncaItattor1 [flow. preaeure, ampe awlngln.gj WAS - OCCURRLNG PRIOR TO ENTRY INTO THIS PROCEDURE. _5. Check If opening )RN-B Traln lB & 2 GO TO Step 8.. SNSWP Supply] STOPPED - CAVITATION.

6. GO TO Emcloe-iire 1 (AIInIng B Train RN to Pond)..

MN LCG OFRN PAGE NO. AR 1kSOD2] 32 oF 111 w tNET 1 Lose oTLow el or RC upplyCraz.soer At] /LX &I i:i :.N 1_I,:. kI tILL.

. CIlc OEE ECEEDD alarm                         IF RM pump on eltiber unit lB I101r ET:l -EiK.                               experiiicIii InIant cavIttlan lnow.

pressure, amps eMngIngJ. THEN I8oiae mon-BelarmIc pIpIng as foLlowe:

CLOSE 1RN-42A iA NDn Es upily [5 o3.

i Hae LinIt2 CLOSE 29N-42A AE No Es upIy 1501).

8. Cliack 1A e-eq incer reet llpht hr.

                                      -              Per1rrn th roILoIng:

a. Fe5et ; i.

i Esetseque1oer. C. IF 1A seuicrrese1 lI9flt Is t, ThEN 2IQ.Ctep .

d. DIsatcii opeiator to open 1 EDA realer iA £G equencer DC cr1rD. Poer.

e. IF ON-AiTraI9 IA & 2A P uppIyi itcr IidIcat1ii s lit, THEN deress PoId l squeicer puFutIn urill IEVDA Ereaer5 is open.

MNC LCSG CF RN PAGE NO. AP1.k550Ci2] ° 111 Case ii Ret fl UNIT 1 Loss D1 Low Level or RC upniy Crossower mn ut/.:*; r.cnu F:;tits:: LL hC L:M &tNEWi 9 Have Unit 2 operator check 2A Have Unit 2 operator pertomt the sequencer reset light LIT.

                            -                            following:

a.. Reset 3I or Un 2.

n. Reset sequeczers :n Unh 1

c. iF 2A sequeicerreselilgllt is .t. ThEN Q TO Step 10.

d. Disoatch operator to open UnIt 2 breaker 2EVDA 5reaker 6 f2A RG Sequcicer DC Control Poweri.

_e IFORN-7A(Traii 1A&2ASNSVdP SLhpplyi switch Jndlcafloi lit THEN deoressani bold 2A sequencer pusFisutton unill 2EVO.k Ereater5 is open. 10 OPEN ORN-7A TraUt 14 & 2A SNSWP Supplyj 11 Watt rp to GO seconds for ORN-74to

PANG LCC CF RN PACE NO. AF.h1.kE.5or,2J & 111 c.is ii R LL%1T I Los olLo.w LeeI or RC tupDIyrer t uIfk 1:lU kIU!*. !3? IJ6)JNtL

12. CIick ORN-7A ITrain 1 & 2.. SNSWP Prfonn the roltowing:

upply -QPN.

                                                      .3  IF AT 1 Nf TihE 1fi.OR 2A RN pump caltate, THEN perrorrn 1heTo(wIrq:

1: if 1A D.G 16 rwirIr, TJjN Inmei lately dispatch oer.3tcco stop 1 A DiG usir.g ernergenc 1rip pLsflbu1tcn. 21 If CA aurnp Is rur.nirg ThEN open 1 A CA pump breaker by resir. TARF an STOP al the Hnie.

R pump.

If uri 2A D I tunnIn, ThEN Irnme<lIately dispatch operatccio stop U iIt2A D:G us:rg eT.rgeicy stop pusl.rn.ittnri

5 If Un 2A CA irip l rur9in, IjiN floury Urt 2 operal or lo DDer 2A DA pump breaker by leresIrg TAftT aid TOP at e same t me. 6, NotIly Jr 1 2 Op!ralor to s!r 2A RN UT 0.

a. [f both oT tIle rciIDAn cntcns are met

All A ralr. rir RN - DEEiEGiZED

                                                     .      1ETA o 2ETA ENERGIZED,
                                                                               -

Th.N swap W urain Eflard RN pcero pcshe ii asfllD:

1) DIsph Opr1Dr to &wp 1EMXM to cpp-iste unIt EjE9cosure 12 i fliTtlig Pcer JLo olle to 1EXH.

MN LC cr 5 ! 111 t%1T 1 f Low LeeI 01 RC .upy Ci.sovr

tuI:.iij I I 1Zl. Ckik CRM<O (Tr&In ID &20 NWP Perform the rolIowIng:

Supplyj ON.

               -
                                                   .             jjç 1DQ3 L RH pirnp ca1ta:es, ]Ii11 peltilIn tie cLow1:

i irm DG irurIrg, ItIH lrrme-lbtely dlaptGh DDertr ID 5tp 1E DiG u5Ir enrgenc op pla8flbL4tDn. Zi j[ ID urp Ii riir 1lrg, fl iLl open lB CApurrpbrkerby deIrflr .DTAt nd OTOP I ttie same tine. 3: 1 RNpun1. i WUni2BDGIsnJn9ngJjN lrrmedbtely dIpth oertr stap JIt2ED,G usng e!rerecy ctp pLLFrnflDn. i iEUr1 cI pump Irur-ln9. IJiEPI ncltTy Ur.t2 0pefiE11o er. Zt CA urnp breaker ty dera5lrg STAR au TCPat tIi sane O Notify Ur1 2 operalor tc 2!: rN pLTD.

i. I.E boi o tie to Ic-* ri cutcns a rie All E trlr, IrC RN .DIyCG DEENE GIZED 1ETE or2ETE - EJERGIZED,

                                                      ]liN swap W lraln shatd 91 OWiO oIpcck 1            llc:

ii DIp1& OFrc O F EM<- tcloppte nItE&Eucoure 13 I:onnIg Poaer il Lo 2EPMHi.

cr i u, zifcttfli ::tf4:;

14. ChecK RN pump on toll units ANY - SOW Stçi 1G.

PUMP SUSPECTED OF BEING AIR BOUND.

15. PerTorm the rolLowLng on both units:

a. Cheek any Unt 1 ar Unit 2 emergency _a. GOTOSteplS.e.

D:G RUNNiNG WLTHOUT RN FLOW

TQ ACGCCIA1tD DIG MX. a.. Cheek flRN-tk :Trair. IAL 2A SNSWP a. GD ID Step 15.d. PPY:I-OPE.

c. CLOSE the id cang waives:

RN-12AC çrrain iA & 2A LLI SLaupiyi

         . JRN-I.3A cr-am 1A & 2A LU Suppx:

1 IT Tedlaie y dSa aLoft operator to stop D:G runrng wihout cooIng waler L419 emergere stca Pushauon. e Check any UnL 1 or Unit2 MD CA e. GOTOSteplS.g. pLTD RUNNING ON RN TRAIN THAT

tt AIR BOUND. t Open pump areaker ny dearessIrg SARV and STOP atlhe same trne.

9. Stop any Lr t I or Ur 2 RN nump thai Is air boLnd.

    $1. Cheek 1 A RN urnp AIR BOUND.
                                  -               Ii. GO TO: Step 15k.

L Check 2RN-7A ;Tralrt IA & 2A ONSWP _L GOTOStep 15_k. SuDpFyi OPEQ

      . Dispawh operator to ient 1 A RN aun:

PER Endosure 7 {1A RN Pump ventngi. Cheek 15 RN unp AIR E.CND.

                                  -               t. G2IQStep 15n.

L CIpeok arr4-9r ;TIdII lb S 2f 3NCWP I 9 J9 oLel) iL.ii. Supplyi .DPEJ.

MN LO CF RN PAGE NO. AFd1.*S5OC2 3 or ill Cas II 9ev 2 TJNIT 1 L0s6 DI Low LiCI or RC uppIy Cros.oer t[ct4:: 14.t c? .;I:1L

15. (cnUnued, DIpach operator to vent 1 E. RN urrip EE Edo6Lre IE RN Purrip Veit ri.

1. Che- Liilt 2ARN pUmp AIR C.LIND.

                                      -                       1. GO 10 step 15.

         . Check ]RN-7A ;Tralr: lAS. 2A ZNAP             . GOTO tep 15.q.

C:uppF) - L DIpmh operator to erit 2A RN urnp PER APi2A55Ef2O iLOS 01 RN], Er ur 7 i:2A RN: Pump Vetn. q Ch& UIt 2 RN prip AIR OLIN0.

                                      -                      q IQ.tep 16.

r. ChE ]RN-E :TrIr 1 & 25 SNO\P 1. 10 step 15.

GuppF OPEN.

6. DIspaich operator to ient 25 RN unp

             £E AP2.Ai55EO Los 01 RN]

Er.ure S i2B R Pump Ve9tn:.

15. flftOUflC& ourmn on page.

r MN AP1.E5OE2 LOG OF RN PAGE NO.

                                                                                                  &1 1 TJ%IT 1                     Ls6 ol L& LeeI or RC .Guppiy Cr.oer                          R I.;(: i F:1J ::L4::

I :kT I NEt;.

17. Jlgrt ATrain RN to thB pond aatoilowa:

a. Chek 39N-7A Trair IA 2A ONGA P a. Perform 1e To

                  - op EN.

1) 11 oera1cc has ee9 dispalchei 10 perform Ercio&ire i ihFtng Power ZLopI.es to 1EMXH,, THEN I.tep 18.

21 J IA D.G Is rL1nir.g. THEN immediately dispatch operator 10 stop IA DiG uslr.9 eniergeny slop puslibulton.

IF 1A cA urnp is rurnlr.q, ThEN open 1A CA pump breaker by deDressirg START and T3F al the same Urrie. 4 lo IA RN pump.

lturl2ADiGis ruriin, ThEN irnme.ilately dispatch ooerattr 10 stop Jilt 2A DG un CTgeicy stop pushouttcn.

6i IF Un 2A CA pump Is rurinin, THEN no1I Urt 2 operaior 10 ooer 2A cA pump breaker by deresIr.9 START aid StOP at the same t me. 7: Noilly Un 12 operalor to stoi 2A RN 8:1 11 dsgrated RN ack-flushi operalor in e.re THEN notify degna1eI operator to ensure 1Ie foIii.ng ive are CLOG ED and reTain CLOSED:

                                                           .      1RN-2 1ARN.traher 5atush Maiua. Guppy ieoi, 3LJX big, 716+, 55-52, room 2 es1 from slrainer 1A1
                                                           .      UnIt 2 aie 2RN-2J 2A RN Z;lraier 5ac11ush Mania upiy iso:i :aLlx bldg. 716+2,CC-53,.

IF ERN-7A is TLi.y cicsei, TIIEN GO TOlep 17.e.

Parent Question (MNS 2009 NRC Exam): Examination Outline Cross-reference: Level RO SRO x Tier# 2 Final Group # 1 K/Mt 076A2.01 Importance Rating 3.7 Service Water Ability to (a) predict the impacts of the following on the (SYTEM) and (b) based on those predictions, use procedures to correct, control, and mitigate the consequences of those abnormal operation: Loss of SWS Proposed Question: SRO 80 1 Pt Unit 1 is operating at 100% RTP with B train components in service with a normal RN system alignment. lB RN pump amps are swinging. The following annunciators are in alarm:

UB RN Pump Suction Lo Press
B RN Pump Discharge Lo Press Which ONE (1) of the following is the correct response to the above conditions based on the implementation of AP-20 (Loss of RN)?

A. Implement Case 1, Loss of Operating RN Train Place the 1A RN pump in service and, Remain on Low Level Intake B. Implement Case 1, Loss of Operating RN Train Swap alignment to the Nuclear Service Water Pond and, Place the 1A RN pump in service C. Implement Case 2, Loss of Low level or RC Supply Crossover Place the lA RN pump in service and, Remain on Low Level Intake D. Implement Case 2, Loss of Low level or RC Supply Crossover Swap alignment to the Nuclear Service Water Pond and, Place the 1A RN pump in service Proposed Answer: D

Explanation (Optional): The swinging amps, lo suction pressure, and lo discharge pressure indicate that the RN pump is cavitating and that the Low Level Intake (LLI) has been lost. Hence Case 2 for Loss of LLI is the appropriate procedure to implement. If the lo suction pressure alarm was not present, it would indicate that the LLI had not been lost and entry into Case 1 Loss of Operating RN Train would be appropriate. lAW Case 2, the Operators will swap to the Standby Nuclear Service Water Pond (SNSWP) and then swap RN pumps. A. Incorrect: See explanation above. Plausible if candidate does not comprehend that the pump suction lo pressure alarm prompts implementation of Case 2. The other indications alone would prompt entry into Case 1. The remaining actions are correct for Case 1. B. Incorrect: See explanation above. Plausible if candidate does not comprehend that the pump suction lo pressure alarm prompts implementation of Case 2.The other actions are reasonable as they would be the correct actions if Case 2 were implemented. C. Incorrect: See explanation above. Plausible because Case 2 is the correct case to implement. The actions are correct if Case 1 had been implemented. D. Correct. Technical Reference(s) AP/1/A/5500/20, Loss of RN, rev. (Attach if not previously provided) 25 pages 2 and 29 35.

(Including version or revision #) Proposed references to be provided to applicants during examination: None Learning Objective: OP-MC-AP-20, Obj. 2 (As available) Question Source: Bank # Modified Bank # NRC Bank (Note changes or attach parent) New Question History: Last NRC Nadeau Retake Exam Exam Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 43.5 Comments: Service Water

Ability to (a) predict the impacts of the following on the (SYTEM) and (b) based on those predictions, use procedures to correct, control, and mitigate the consequences of those abnormal operation: Loss of SWS KA is matched because the candidate must evaluate the provided conditions to determine the impact on the system and determine from those indications which is the appropriate procedural strategy. This question is an SRO Only question linked to 10CFR55.43(b)(5) (Procedures) because the question can NOT be answered by knowing systems knowledge alone, it can NOT be answered by knowing immediate actions from AP-20, and it can NOT be answered by knowing AP-20 entry conditions alone. It DOES REQUIRE the candidate to recall the AP-20 mitigating strategy and specific procedure steps within the body of AP-20 to be able to correctly answer the question. Modification to stem in Atlanta.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 80 SYSO63 2.1.23 - DC Electrical Distribution System YS063 GENERIC Ability to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10 / 43.5/45.2/45.6) With both Units at 100% RTP the following occurs:

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 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 AVCIYC Chiller within 37.5 minutes. D. Complete 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 hour and 15 minutes. Tuesday, July 13, 2010 Page 232 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 80 D58o eneraI 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 aboveBatteiy charger EVCS would only be utilized if it was aligned to the battery prior to the event. This woui 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 seernalogical 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 ifaVC chiller fails is 37.5 minutes. Answer D Discussion - INCORRECT: See explanation above. See explanation above.Train A VCIYC power and water only need to be swapped if a station blackout has occurred and that both DIGs 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 1 hour and 15 minutes from the time of the blackout. asis for meeting the KA JA 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 ofnot 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 0311112010) 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.

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. Job Level Cognitive Level QuestionType Question Source SRO Comprehension BANK MNS Bank NRC QIOO [eIopment References 1ent References Provided P107, Loss of Electrical Power (Rev 28) page 9

  -07, Loss of Electrical Power Bkgd (Rev 7) i-gsl4&15 OP-MC-AP-07 Obj: 2 Tuesday, July 13, 2010                                                                                         Page 233 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 80 SYSO63 2.1.23 DC Electrical Distribution System

1S063 GENERIC

.oility to perform specific system and integrated plant procedures during all modes of plant operation. (CFR: 41.10/43.5 /45.2/45.6) 401-9 Comments:                                                                  RemarkslStatus
                                                                               ---_______

401-9 Comments: No comment. Resolution / Comments: N/A Tuesday, July 13, 2010 Page 234 of 294

Question 80

References:

OP-MC-AP-07 Obj: 3 OBJECTIVES NNLLL OBJECTIVE L L P P 0 OORSR ROO 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 AP/! /A/5500107 9 of 395 Case I UMT 1 Loss of Nomial Poter to Both IETA and 1ETB Rev. 28 Ac::ON;EXPEOTED PESPONSE RESrONE: NOT OBTAINED CAUTIQN If operating train of VCIYC has failed, it may be time critical to swap operating VCIYC trains.

17. Initiate EPI1/N5000IG-1 (Generic Enclosures), Enclosure 13 (VC And VA System Operation) within 30 minutes of B/O.

18. Check SIG Pressures STABLE OR

M 4jftflME SJG pressure goes down GOING UP. in an uncontrolled manner AND reactor tripped, THEN CLOSE the following valves:

All MSIVs
All MSIV bypass valves.

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. API1INSSOOI07 10 of 395 Case i UNIT 1 Loss of Nom,aI Por to Both 1 ETA and I ETB Rev. 28 AI::CN;EXPICTED RESPONSE Rzs;ONS: NOT OBTAINED

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.                                             GQJQStep2D.

b. Main Turbine EMERG BRG OIL K Start pump.

PUMP.

c. DC B/U VAP EXTRACTOR. c. Start vapor extractor.

d. 7k CF PUMP TURB EBOP. d. Start pump.

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

f. Check OAC AVAILABLE.

                          -                                  f. Perfomi the following:

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

2) Observe Caution prior to Step 20 and Q IQ Step 20.

g. Check computer point M1D0581 (Ui g. Dispatch operator to ensure Unit 1 HAIR Gen Air Side Seal Oil Backup Pump) - SIDE BACKUP pump running (Unit I ON. Turbine Bldg. 760, 1 F-23).

CAUT1QN 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 B!O.

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 1 DIG), it must be swapped to the other unit within an hour (MCC-1 381.05-000-0220, 1 25VDC 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 VC/YC Chiller within 37.5 minutes. D. Implement Generic Enc. 13 (VC and VA System Operation) to swap Train A VCIYC Chiller power and water to Unit 2 and restart chiller within 1 hour and 15 minutes. Proposed Answer: A

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 81 SYSO 15 A2.O 1 Nuclear Instrumentation System (NIS)

bility to (a) predict the impacts of the following malfunctions or operations on the NIS; and (b based on those predictions, use procedures to .orrect, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 /43.5 /45.3 /45.5) Power 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 hours

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 OR movable incore detectors

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

2. AXIAL Tuesday, July 13, 2010 Page 235 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 81 L2581 General 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: 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. [12_is correct. Answer B Discussion

.RECT:_See explanation above.

Answer C Discussion ITh4CORRECT. 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 INCORRECT. See explanation above. AUSIBLE: 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 [it eexceeding AXIAL distribution limits. Basis for meeting the KA iKA is matched because a power supply failure for an NIS channel has occurred and the applicant must determine the impact on Quadrant [er_Tilt determination in accordance with Tech Spec requirements. Basis for Hi Cog E___ 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 hour 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 hour 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 jech Spec basis document and in the surveillance requirements. Job Level Cognitive Level QuestionType Question Source SRO Memory NEW

   )velopment References                                                                      ient References Provided Learning Objective:

1) IC-ENB #19

References:

Tuesday, July 13, 2010 Page 236 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 81 2581

1) Technical Specification 3.2.4 QPTR Technical Specification 3.2.4 Basis YSO15 A2.O1 Nuclear Instrumentation System (NIS)

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 /45.5) Power supply loss or erratic operation 401-9 Comments: -- RemarkslStatus 401-9 Comments: No comment. Resolution / Comments: N/A Tuesday, July 13, 2010 Page 237 of 294

Question 81

References:

From Tech Spec 32.4 QPTR: QPTR 3,2.4 SURVEILLANCE_REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1 --------r I

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 hours and every 12 hours 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 POWER>75% RTP. Verify QPTR is within limit using the movable incore 12 hours detectors. McGuire Units 1 and 2 3.2.4-4 Amendment Nos. 184/166

From Tech Spec 3.2.4 Basis: 330 PTR 632.4 B 3.2 POWER DISTRISUTION LIMITS B 3.2.4 QUADRANT POWER TILT RATIO (OPTP) BASES BACKGROUND The OPTR limit ensures that the qross radia power distributon remains consistent vith the design values used in the safety analyses. Precise radial power distrdxition measurements are iiade during startup testing, after refueling, and perodically dtring power operation The power censity at any pant in the ore nust ba limited so that the fuel design criteria are maintainod. Together, LCO 3.2.3, AXIAL FLUX DIFFERENCE (AFD). LCO 3.2.1, and LCO 3.1.6 Control Rod Insertion Liiiits, pro ide lini its on process ariables that cfraraclerize and control 9 the thrøe dimensional pnner diflrihiitirn of the reactor -nre Control of these variables ensures that the core operates within the fuel desn criteria and that the power d stribution remaiis within the bounds used in the safety analyses. APPI CARl F This I CO pracltidas core po.ver ttrihiihon that winlatR the foIloving SAFET ANALYSES fuel design criteria:

a. During a large break loss of coolant accident (LOCA). there must he a high level o probability that the ak dadding temperature coes not eceed 2200°F (Ref. 1)

b. The IDNBR calciiated for the hottest fuel rod in the core must be above the approied DNBR limit (The LCO alone is not sufficient to preclude DM6 criteria violations for celain accidents, i.e.. aacidents in which the event itself chanqesthe core pnerdistributior. For these events, additional checks are nude in the :ore reloac desijn process against :he permissible statepoint power distr butioas.);

c. During an ejected rod accident, the energy deposition to the fuel nust not exceed 280 cal/gm (Ref. 2); and

d. The control rods must be capable of sautting down the reactor wth a minimum requied 5DM ith the highest worth contol rod stuck fully wthdrawn (Ret. ).

The LCO linits on the AFD, the QPTR. the Heat Flux Hot Channel Factor (XX.Z)), the Nuclear Enthalpy Rise Hot Channel Factor (F(X.Y)), 0 (F and control bank insertion are established to precude core power detributions that exceed the safety analyses limits. McGuire Units 1 and 2 B 3.24-1 Revision No. 10

QPTR B 3.2.4 DA3Cf3 AClONS (cortinuec) reaching RTP. As an added precaution, lithe core poser does not reach RTP Mthin 24 hours, but is increased slowly, then the peaking fac:or surveillances must be performed wilhin 4E hours of the time when the more restrictive of the pwer level limit derermined by Requiecl Action A.1 o Al is exceeded. These Conipletioi Times are intenced to allow adequate time to ncrea;e THERMAL POWER ta above the more restrictive iinit of equired Action A. 1 or P.2, while not permitting the core to remain cith unconfirmed power distributions far extended periods of time. Required A;Liuri: Al is itudilied by i Nuieilii[sta1s lIij1 (tie e.4iiiy factor surveillances must be done after the exccre detectors have been calibrated to shos zero tilt (i.e., ReqLlired .ction &. The intent of this Note s to have the peaking factor surveillances perfonied operating power levels, which can only be accomplished after the excore delectors are calibrated to show zero tilt and the core rett.rned to power. Si If Required Actiors A. 1 through &7 are not conpleted within their associated Competion Times, the unit must be brought to a MODE or condition in which the requirements do not apply. To achieve this status, THERMAl. DOWER must be reduced to 50% RTP within 4 hours. The allowed Completion Tine of 4 hours is r&sonable, based on operating experience regarding the amcunt of time required to reach the reduced power level without challengirg plant systems. SURVEILLANCE SR 3.2.4.1 REQUIREMENTS SR 3.2.4.1 is mocitied by two Notea Note 1 allows UPIR to be calcu ated with three power range ciannels if THERMAL POWER is < 75% RTP and the inputfrom one Poner range Neutron Flux charnel is inoperable. Note 2 alloys performance of SR 32.4.2 in lieu of SR 3.2.4.1 if more than one input from Power Range Net. tron Flux charnels are inoperable. This Surveillance verifies that the CPTR, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits. The Frequency of 7 days when the QPfl alarn is OPER \BLE io acccptable 1 because of the low probability that this alam can remain inoperable withoit detection, McSuire Units 1 and 2 b 3.2.4-b Revision No. 10

QPTR 53.2.4 BASES SURVEILLANCE REQUIREMENTS (continued) When the QPTR alarm is inoperable, the Frequency is increased to 12 hours. This Frequency is adequate to detect any relatively slow changes in QPTR, because for those causes of QPT that occur quickly (e.g., a dropped rod). there typically are other indications of abnomialitv that prompt a verification of core power tilt. SR 3.2.42 This Surveillance is modified by a Note, which states that it is required only when the input from one or more Power Range Neutron Flux channels are inoperable and the THERMAL POWER is 75% RTP. With an NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes deqraded. Large tilts are likely detected th the remaining channels, but the capabiliw for detection of small power tilts in some quadrants is decreased. Perfomiing SR 3.2.4.2 at a Frequency of 12 hours proides an accurate alternative means for ensuring that any tilt remains thin its limits. For purposes of monitoring the QPTR when one power range channel is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the indicated QPTR and any previous data indicating a tilt. The incore detector monitoring is perfomwd with a full incore flux map or two sets of four thimble locations with quarter core symmetry. The tto 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 tilt This can be compared to a reference smmetric thimble tilt, from the most recent full core flux map, to generate an incore tilt. Therefore. incore tilt can be used to confirm that QPTR 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, which 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 3.2.4-6 Revision No. 10

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 82 SYSO41 2.4.11 Steam Dump System (SDS)/Turbine Bypass Control

YSO41 GENERIC now1edge of abnomial condition procedures. (CFR: 41.10/43.5/45.13) Given the following conditions on Unit 1:

The unit is operating at 108 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 Tuesday, July 13, 2010 Page 238 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 82 2582

.3eneral Discussion In 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. The 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 NI would be adversely affected and if temperature were allowed to fall 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 551°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 I 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. 9 L AUSIBLE: 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. 2 is correct and therefore plausible. Answer D Discussion CORRECT: See explanation above. Basis for meeting the KA or this sencario, the applicant is given a malfunction of the Steam Dump System and is asked to demonstrate a knowledge of abnormal procedure 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 This 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 SRO knowledge.

Job Level Cognitive Level QuestionType Question Source SRO Comprehension BANK BANK Q 491 Tuesday, July 13, 2010 Page 239 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 82 eveIopment References Student References Provided

 -0l (Rev 16) page (7 of42)

IS 34.2 Basis SYSO4I 2.4.11 Steam Dump System (SDS)/Turbine Bypass Control

SYSO41 GENERIC Knowledge of abnormal condition procedures. (CFR; 41.10 / 43.5 / 45.13) 401-9 Comments: - RemarkslStatus 401-9 Comments; No comment. Resolution / Comments; N/A Tuesday, July 13, 2010 Page 240 of 294

Question 82

References:

From AP-Ol: MNS STEAW LEAK PAGE NO. API1 IAISSOD?0i 3 of 42 UNIT 1 Rev. 16 AcI:ct1/ExrZCTED aEsrcnsE REsrtTsE 1OT OSTAnaD C. Operator Actions

1. Monitor Foldout page.

2. Reduce turbine load to maintain the following:

        . Excore Nis LESS ThAN OR EQUAL
                           -

TO 100%

        . NC Loop DRs LESS THAN GWF D;T
                              -
        . T-Avg-ATT-REF.

3. Check containnient entry IN

                                     -                .

Q TQ Step 5. PROGRESS F:

4. Check steam leak KNOWN TO BE

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

a. Announce AlI personnel evacuate Unit 1 containmenr.

b. Actuate the containment evacuation al ami.

c. REFER TO RPIUIA/5700101 1 (Conducting a Site Assembly, Site Evacuation, or Containment Evacuation) as time allows.

5. Check Pa 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 (Sll Actions For Steam Break In Modes 3 and 4).

MNS STEAM LEAK PAGE NO. AP/11A15500101 4 of 42 UNIT I Rev. 16 A:TIoN/E::PETEr .ESCNEE EPONSE OT O3TAED

6. Check Pzr level STABLE OR GOING UP.

ls requi

a. afii cngiiq fic less t 20C 3Pi t 11 t flies in subs oentia1 alternate pS.

f_owpa+/-.

b. ure

ENI

c. 5PEN INV-2Z1 :Ufl.aterli FIov Dorzrol : ile maintaininc NC:

pump sea: flow geatert-ian 6Pi..1.

d. ce or ioiatetetc

e. acdit.ora NV pi

f. Fzr leeI qoic do vith aim.

arging f.o:. Th4 QII

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

_8. QjStep12.

MNS STEAM LEAK PAGE NO. AP/ 1 /A/5500/O 1 5 of 42 UNIT 1 Rev. 16 AcTIc/ExPECTEt ESPDNSE ESCNSE lOT QBTA.EE

9. unit Perform the following:

NOTE If reactor trip breakers

       . f          prior to                        are closed and T-Ag is less than 553F, a feedwater isolation will occur in the next step.

a. Open reactor trip breakers.

b. CLOSE all MSIVs using individual aIve P-tenti.J. a..rnate pushbuttons.

fl.:wpath.

c. IE EPi A/5000ES-O. 1 (Reactor Trip Response) has been implemented, THEN perform the follo.ing:

1) THROTTLE SIG feerl flow to:

                                                  . Minimize cooldown
                                              .      Maintain at least one S/G NP leel greater than 11%.

_2) GOTO Step It

d. IF feedwater isolation has occurred AND SIG Levels going down in uncontrolled manner THEN perform the following:

1) StartCApump(s).

2) WHEN desired to feed SIGs with CMICF, ThEN REFER TO API 1 1A1550C?06 SIG Feedwater Malfunction

e. THRO1TLE S/G feed flow to:

                                              . Minimize cooldown

                                              . Maintain at least one S!G N/R levet greaterthan 11%.

_f QjQStepi1.

ACTION EZPECTEO ESPDNSE PESiONSE NOT OBTAflZD

10. FbSé&toran .lVsasfoll Pctentiaj. alternate

_a. flcwpath. b.

c. Sntinue wit

_cI. rQIQEP/1!A. or Safety lije Ii. jfAIANYIiMEPzrlevelgoesbelow4% AND cannot be restored using normal charging, THEN perform the following:

a. Ensure reactor tripped.

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

c. Check Pzr pressure prior to event - c. Q JQ Enclosure 2 (S/I Actions For GREATER THAN P-i i (1955 PSlG. Steam Break In Modes 3 and 4).

d. GO TO EP?i.IAI5000IE-0 (Reactor Trip or Safety Injection).

12. Announce occurrence on paging system.

ACTION.!EXPECTED PESrONSE RESPONSE NOT DETAilED

13. Identify and isolate leak on Unit 1 as foIlows

a. Check SM PORVs CLOSED.

                          -               a. IF SIG pressure is less than 1092 PS 1G.

THEN perform the following:

1) CLOSE affected SJG SM PORV manual loader.

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

a) CLOSE SM P0kV isolation valve. b) IF SM P0kV isolation valve still open, ThEN dispatch operator to CLOSE SM P0kV isolation valve.

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

CLOSED. THEN perform the following:

1) SelecttFFRESETonthe following switches:

STEAM DUMP INTLK BYPASS CHANNEL K
STEAM DUMP INTLI< 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 Q repaired. THEN return the following switches to

STEAM DUMP INTLK BYPASS CHANNEL A

                                             .      STEAM DUMP INTLK BYPASS CHANNEL B.

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 imply a recriticallity concern F irst consideration is moderator temperature coefficient (MT 3.1.3, Moderator Temperature Coefficient (MTC). In the ccident analyses, the MTC is assumed to be in a range frc ive to negative and the operating temperature is assumed to the nominal operating envelope while the reactor is critical. minimum temperature for criticality helps ensure the plai ated consistent with these assumptions. The second consideration is the protective instrumentation. Because certain protective instrumentation (e.g., excore neutron detectors) can be is the correct answer affected by moderator temperature, a temperature value within the ILed 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 (DBAs), 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). LCD Compliance with the LCD 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 LCD 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, LCD 3.4.2 is applicable since the reactor can only be critical (keff 1 .0) in these MODES. The special test exception of LCD 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 LCD. 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 LCD.

Parent Question (Bank Question 491): 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. ANSWER: C

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 83 2583 SYSOO2 A2.02 Reactor Coolant System (RCS)

bility to (a) predict the impacts of the following malfunctions or operations on the RCS; and (b) based on those predictions, use procedures to .,orrect, control, or mitigate the consequences of those malfunctions or operations: (CFR: 41.5 /43.5 /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
1NC-32B (PZR PORV) fails open
AP-il (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 In accordance with AP-Il, the crew shall wait until the Reactor is tripped (1)

The crew will subsequently transition to (2) Which ONE (1) of the following completes the statements above? A. 1. AND then initiate SI, even if the low pressure SI setpoint is exceeded

2. AP-34 (Shutdown LOCA)

B. 1. AND then initiate SI, even if the low pressure SI setpoint is exceeded

2. E-O (Reactor Trip or Safety Injection)

C. 1 Q the low pressure SI setpoint is reached to initiate SI

2. AP-34 (Shutdown LOCA)

D. 1. OR the low pressure SI setpoint is reached to initiate SI

2. E-O (Reactor Trip or Safety Injection)

Tuesday, July 13, 2010 Page 241 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 83 General Discussion 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- 11 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 NCORRECT: See explanation above. LAUSIBLE: Part 1 is correct and therefore plausible. 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 D Discussion LC011ECT 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 I dated 03/11/2010 for screening questions linked to 10CFR55.43(b)(5) (Assessment and selection of procedures): I) 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 tions.

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

 .jnditions 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- Ii. 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 Tuesday, July 13, 2010 Page 242 of 294

__ ____ FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 83 E2583 the same step directs transition to E-0 if the unit is in Mode3or above with the CLAs not isolated. Therefore, this is SRO levelknowled ge. Job LeveT Cognitive LeviQuestionType Question Source en s t lisio h C nw Development References Student References Provided Learning Objectives:

1) N/A

References:

1) AP-l 1, Pressurizer Pressure Anomalies SYSOO2 A2.02 Reactor Coolant System (RCS)

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

                                                                                                          / 45.3 / 45.5)

Loss of coolant pressure 401-9 Comments: RemarkslStatus 40 1-9 Comments: Put The crew shall wait until the reactor is tripped in the stem and start each choice with AND / OR as indicated. It will read much better. Resolution / Comments: Revised question but in order to make this work as suggested had to make this two separate fill-in-the-blank questions, See edfor proposed revision. Tuesday, July 13, 2010 Page 243 of 294

Question 83

References:

From AP-1 I (Pressurizer Pressure Anomalies): MNS PRESSURIZER PRESSURE ANOMALIES PACE ND. API lAdS 600111 2 &t 9 UNIT 1 Rev. 10 ACIION.EEPZCTEJ PZSrONEE P1ErCNSE ;D7 OST.A:NSD S. Symptoms

Pu pressure channel failed
Pzr pressure going down in an uncontrolled manner
Pu pressure going up in an uncontrolled manner o Any Pzr PORV or spray valve failed open
PZR PORV DISCH HI TEMP alarm
PRT HI TEMP alarm.

C. Orator Actions Check actual Pzr pressure HAS GONE

Q JQ Step 17. DOWN.

      .1)      Check all Par pressure channels INDICATING THE SAME.
                                               -

liE either controlling channel is malfunctioning, I[IE!4 place PZR PRESS CNTRL SELECT switch to backup channel. Check Pzr PORVs CLOSED.

                                 -                           Perform the following:

a. Close PORVs.

b. IF PORV xiii not close, THEN close P0kV isolation valve.

ED Check Pzr spray valves CLOSED.

Perform the following:

a. Close Pzr spray valve(s).

I

b. IF AT ANY TIME a reactor uip occurs AND spraj valve still open. THEN stop 1A and IS NC pumps.

MNS PRESSIJRJZEfl flPESSWE ANOMlJES PACE ND. APLI ILJE5DCl 1 3 of 9 UNiT 1 Re,. 10 A;I:Ou.. E:ncTED PL3;cIsE ECTSE NOT c3TA:xE:

5. Cliec Fzr PCRVs -CLOSED. Perform [he following:

a. II- associated PURV solaton valve wU rioi cIus AND pwsstiw yuiny duwu rapidl. ThEN:

1) iF in Mcde 3 or above, prior to CLA sola:on. THEN:

a) Trip reactor. b) WHEN reactor tippedORaLto 5/I etpoint reached. THEN esure Sl initialed. C) p TO EPI1AISOCO/E-O (Reactor Trip or Safeti Injection). II- n Mccc alter C. h->- 2) 1odespr

b. Close asociuLed PORV iiiletJiuiii valve acfoIlo:

                                                  . F INC-323 (P PORV) filled, THEN close INC-271 (PZR PDRV Drri Isol For 1 f\C-32B).
                                                  . F INC-34A (PZP P0kV) failed, THEN dose 1 NC-71) (P7P PE)PV
                                                         )rn sal For it%C-MA).
                                                  . F INC-365 (PZR PORY) failed, fl lEN close I NC-269 (PZR PDRV Drn Isol For INC-3B).

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 84 2584 APEOI5/017 AA2.02 Reactor Coolant Pump (RCP) Malfunctions

( ility to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): (CFR 43.5 /45.13)

  - bnormalities in RCP air vent flow paths and/or oil cooling system 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
1C 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. C. Reduce reactor power to < 10% using AP-04 (Rapid Downpower), then stop the 1C NCP. D. Reduce reactor power to < 5% using AP-04 (Rapid Downpower), then stop the IC NCP. Tuesday, July 13, 2010 Page 244 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 84 258 neral Discussion In accordance with AP-08, if an NC pump must be stopped reactor must be tripped if the unit is operating in Mode I 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 OP/l/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). It is also plausible for the applicant to believe that the reactor would be at greater than 5% after a trip since initial decay heat load immediately after a trip is approximately 7%. Answer B Discussion CORRECT. See explanation above. Answer C Discussion INCORRECT. See explanation above. PLAUSIBLE: This answer is plausible because stopping the NCP when power is less than 10% is reasonable since the NC low flow trips are defeated when less than 10% power (P-b). - Answer D Discussion

C0T See explanation above.

PLAUSIBLE: This answer is plausible because the NC pump is stopped when power is less than 5%. However, the reactor is tripped instead of [pforming a Rapid Downpower.

  -Basis for meeting the KA                                                                                                                        --

( \ e KA is matched because an malfunction of the oil cooling system has occurred and the applicant must determine the appropriate AP-08

      ,tions 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 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 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. ______________

   }evel                     Cognitive Level           QuestionType                                     Question Source SRO                   Memory                      NEW Development References                                                                         Student References Provided I____________

Objectives: eferences: j)P-08, Malfunction of NC Pump Tuesday, July 13, 2010 Page 245 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 84 APEO15/017 AA2.02 Reactor Coolant Pump (RCP) Malfunctions

ility to determine and interpret the following as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow): (CFR 43.5 /45.13) bnormalities in RCP air vent flow paths and/or oil cooling system 401-9 Comments: RemarksiStatus 401-9 Comments: No highest power level is listed in distractor C, therefore C is NP. If there is no ATWS, why would power be 10%? Distractor A is NP. Nothing in the reference supports 10%. The Q is U

                                                                               *because of 2 NP distractors.

Resolution / Comments: Revised two of the distracters to eliminate potential NP distracters. See attached document for proposed fix. Tuesday, July 13, 2010 Page 246 of 294

Question 84

References:

From AP-08: MNS MALFUNCTION OF NC PUMP PAGE NO. API1!A/5500/08 14 of 24 Case II Rev. 12 UNIT 1 NC Pump Motor or Motor Bearing Malfunction ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED B. Symptoms

NC pump stator winding temperature going up
NC pump motor bearing temperatures going up
NC pump upperIlower oil reservoir level computer alarm.

C. Qperator Actions

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 ifi trip criteria valid, ]IIEN Q TO Step 5.

operating limits:

              . All NC pump stator winding temperatures
                 - LESS THAN 311°F

              . All NC pump motor bearing temperatures
                 - LESS THAN 195°F

              . All NC pump oil reservoir level computer points INDICATING BETWEEN
                         -

(-)1 .25 AND (+)1 .25.

3. IF AT ANY TIME any operating limit in Step 2 exceeded, THEN GO TO Step 5.

_4. GOIStep6.

MNS MALFUNCTION OF NC PUMP PAGE NO. AP1iA55OO/O8 15 of 24 Case II UNIT 1 NC Pump Motor or Motor 6earing Malfunction Rev. 2 ACTION/EXTECTED RESPONSE RESPONSE NOT ONTAINED

5. Stop affected NC pump as follows:

a. IF A or 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).

b. Check unit status- IN F1ODE I OR2 b. Perform the following:

1) Stop the affected pump.

2) IF at NC pumps are off, THEN perform the following:

a) Secure any boron dilution in progress. b) If in Mode 3, TH.E4 immediately open Reactor Trip Breakers C) !E the step above results in rods dropping AND Pzr pressure is above P-li. THEN GOTO EP1iAf5OOOE-O (Reactor Trip or Safety Injection). _3) GOTOStep6. c Trip reactor d WHEN reactor power less than b, THEN stop affected NC pump. e GO TO EP/11A15000/E-O (Reactor Trip or Safety Injection).

6. Announce occurrence on paging system.

MNS MALFUNCTION OF NC PUMP PAGE NO. APR /AS5OD/OB Case II 17 of 24 NC Pump Motor a Motor Bearing Malfunction Rev. 12 UNIT 1 acIIcwE:Ps:TEr RESCNSL FLEPONSE 10t OBTA:NED Perform the following: a. Stator windirg temperalures (OAC) STABLE OR GOING DOWN

       . v1otor bearir ternreratires                      b. SCpurp needs t53 SflBLE OR C INC DC I                                    .....pefcrri tre fojj a Vibration NDRMAI.
                  -                                           1)              III    !Riii ii
    . Id reseroirle  els (ZAG) ST?B

                                                                   . CR1 LNG 100/003 Ccntrclling Prcueduie Fur Unit Ope[aLioir).

Fncinsnre 4 2 (Pn.wr ReHririinn) Distracter OR Pausibility _.AP 4 Do4

2) WI-lEN in Mode 3. 4, or 5, THEN perform the following:

a NC pumps need to be stopped, ifiEN perfomi fle to lowing: (1) Secure boron lution. (21 Do rot continue until rods inserted and reactor trip brothers open. b) IF A cr B NC pump is the afectcd punp, THEN CLOSE associated spray iabe:

C iA NC Loop ZR 7

1NC-2 Spray Control:

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

c) Stop affectec NC pt.nip. ii. Check NC pumps ANY RUNNING.

                          -                              IF boih ND pumps off AND no EP in effect Tll[N REFER TO APIIjA1550(IO9 (Natural Circulation) as tine allows.

FflD

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 85 2585 APEO22 2.4.46 Loss of Reactor Coolant Makeup

                   -
\PE022 GENERIC Ability to verifi that the alarms are consistent with the plant conditions. (CFR: 41.10/43.5/45.3 / 45.12)

Unit 1 is operating at 100% RTP when the following alarms are received:

1AD-7 / Ji (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 byAP-12?

2. What actions are directed by AP-1 2 regarding the restoration of letdown during the subsequent recovery?

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

2. Pressurize the letdown system locally.

B. 1. Close I NV-IA, 2A (NC LID Isol To Regen Hx) and ensure that the Letdown Orifice Isolations (INV-458A, 457A, 35A) auto-close.

2. Pressurize the letdown system locally.

C. 1. FIRST close the Letdown Orifice Isolations (1 NV-458A, 457A, 35A) and then close 1NV-1A, 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 L/D Isol To Regen Hx) and ensure that the Letdown Orifice Isolations (INV-458A, 457A, 35A) auto-close.

2. Pressurize the letdown system from the Control Room.

Tuesday, July 13, 2010 Page 247 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 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 Hx Letdown Hi Temp alarm is in, they should close the isolations to the Regen Hx (NV-IA, 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-lA 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 (1AD7-12) 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. ethe 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-1A and 2A which states: IF AT ANY TIME, REGEN HX 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 verify 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. sis for SRO only his 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) This question can NOT be answered by knowing systems knowledge alone. This requires tha applicant to analyze a given set of alarms and determine what plant conditions could have caused that combination of alarms. The applicant must then determine what procedural actions from Tuesday, July 13, 2010 Page 248 of 294

___ ________ FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 85 2585 AP-l2 are appropriate for plant conditions. This question can NOT be answered by knowing immediate operator actions. AP-l2 has no immediate actions. 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 frorn Job Level Cognitive Level QuestionType Question Source SRO Comprehension NEW Development References Student References Provided Learning Objectives:

l)AP12002

References:

1) AP-12. Loss of Letdown, Charging. or Seal Injection APEO22 2.4.46 Loss of Reactor Coolant Makeup

APEO22 GENERIC Ability to veri1 that the alarms are consistent with the plant conditions. (CFR: 41.10/43.5 /45.3 /45.12) 401-9 Comments: - Remarks/Status 401-9 Comments:

                                                                                 # 2 in the stem can be deleted. This Q is asked in stem #1. Or you can state in #2: What are the restorative actions. As written, the stem seems a bit convoluted.

Change BI and Dl to ensure letdown orifice isolation valves INV XXX auto close. The way it currently reads can be construed as teaching even though it is not true. Resolution / Comments: Revised question as suggested by Lead Examiner. See attached document for proposed fix. - Tuesday, July 13, 2010 Page 249 of 294

Question 85

References:

From AP-12 (Loss of Letdown, Charging, or Seal Injection): MNS LOSS QZ LETDOWN, CHIsRGING OR SEA_ INJECTION PAGE JO. AP/1IA55001i2 3of 4) UNIT 1 Rtv. 22 z:==:DDZ:Z:C1cDDflazzz C. Operator Actions

1. Check iIctidiqir.q is dlkJIIetI to L fit u1ciiiqinq ti.uuqli Recene:aiive Regenera:Ne -lx LS follcms: Hx has occurred. ThLN perform me i tolowing:

I tharcinq how URtA I ER I HRN 20

(3PM a. Ensire th the foiIoinq va[ves are Ci OSFD z INV241 U Seal Water nj FFo. Contri) THRCflLED OPEN

                         -                                     _:      NV-I LEA (7E (3PM LiD Orfice Cut[ct Cont ioif z INV-244A (Charging Line Cant Outside isoli OPEN
                     -                                            : *W-457A (45 GPM LiD 0 rice Outlet Cont Isol)

_: 1NV-2455 fChargiiig Line Cant Outside lsoij OPEN.

                     -          .                                 :  NV-35A \anab1e LID Crifice (ullel N                                   CcntIsl) h   ir IW-7A (JC I ID isol Tn Regen Hx) lie the applicant is not given            closed, THEN CC) tO Step I tiese spectic hidicatlors, they IT M              2ECEN H)( LETON should be abe to determine                       TEMP ian-n (lAD-7           is it torn the contination ol arrns
                                                                                            -

CLDSE tre foiIo0.ing valves: flat charging flow has been lost.

NV-IA(NCUbisotiokegenHx)

NV-2A (MC Lt sot To Regen Hx).

Letdown Recovery Actions from AP-12: MNS LOSS OF LETDOWN. CHAPG NO OP SEAL INJECTION PAGE NO. APLI/iJESOO)12 20 of .13 UNET] L2. CIted the lulluwitig aites - OPEN; Pei knit Lht Iollunirig; _: I NV-IA (NC LID Iso To Regen -b) a. IF nornlGl letcown kriownto be unavailable. JU fl Step 49. z I NV-2 (NC LiD Isol To Regen H).

b. Pior to openng 1 NV-lA ar 1 NV-2A in subsequert step. enawe that all personnel are ot of lower cortainnient

(çotentki water hammer even:).

c. Observe Cau:ioi prior to Step 44 and GO 10 Step 44 L3. GOtQStep.t8.

CAUTION Establishing normal letdown without ocal pressurizction may anise some water hammer. Afternte 1$ Dei niiiie ilut,,ditioiis allow iiiiiiiediate Path its tot ii lion of 1101 iiiul letdown im 1uIIovs;

a. Check boil-i 1NV-1A (IC L/DIwlTo Nr. a. [QjwF<NCbr Stp-4.d.

Reqen Mx and INV-2A (MD L/D sol TcC

Regen Hx) OPEN WITHI1 THE LAST

dli MINJ lhS.

b. Check orifice isoLation vaNes AUTO - b. Perform the following:

CLOSED. 1 IF orth i3daon valves were manually Dlcsed while both iN!- IA

                                                                          ,ind Nv/-?A              T}-IFN GO rnSteaLS.

2; Q TO Rfff for Step 44.cl.

MNS LOSS OF LETDOWN. CHARGING OR SEAL INJECTION PAGE NO. AP?IINSSOW12 21 of 43 TENIT 1 Rev. 22

44. (Continued)

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

valve went closed on the OAC by THEN Q TQ RNQ for Step 44.d. performing the following:

1) Enter turn on code ARCHIVE.

2) Ensure OAC automatically populates START TIME and STOP TIME. (previous hour).

3) Enter group name AP12.

4) Click FS VIEW PID.

d. 1 r mrr 1 d. Perform the following:

1) IF excess letdown is in service.

THEN observe Caution and Note prior to Step 46 and G.QTQ Step 46. NQTE Establishing normal I letdown requires local pressurization of letdown header. Since this action takes significant time, establishing excess letdown first may be desired. _2) lFATANYTlMEitisdesiredto establish excess letdown, THEN Q JQ Step 49.

3) Observe Caution and Note prior to Step 46 and QQIQ Step 46.

45. *JQStep*

MNS LOSS OF LrDDWN, CHARGING OR SEAL INJEDTIO1 PAGE NO. APiiq55GO?l2 22 of 43 flWf 1 CAU11ON It is preferable to locally essurize (he letdown line prior lo establishing letdown t1 due to possthle water hwnmei. NOTE If olant conditions requie minediate restoration of nci-nal letdown OSIcI may waive the requirement lo locally pressurize We tdon ieacter. 4 If the applicant ccncludes that dosing the

46. jf normal letdwvn is required prior to Orfioe Isolation Valves first is the correct kcally,pressudzing letdown herider HFN JQ. Step IS. response or does not recall wie the letdovgri line shuld be pressurized localy irelative to the :lcst re of the Letdown 47 ____I Iscialions arrt the I tdnwn Orifice lsclatinns) they wiild erri up at this step b.

c. GO TO Step 48.

e. I.

MNS LOSS OF LEIDOAN. OHARC NC OR SEAL INJECTION PACE NO. ARi/A55Ol2 240:43 UNIT I Re 22 4R Fsrahlish normal lctdnwn as follows: GO TO Step 49. hnsure iNVb9 i U I Variable LU \-hs is the beçinning of The sequence o Orifice Octict Flc Cntr) is CLOSED. \ steps to pressiirve letdown from the Control oom. You end up here

b. Place I NV-I 24 (Letdo*n Pressure regardless DT how long letdown as Coitrcl) in manual bet*een 1C-20% isolated provided 1NJ-iA & 2A cosed OPFN riorto the orilice iscIatin vakeE.

c. CheckthetoLowng qalves- U-tN: Th. hnsure all rerwnrei are outotbwer contain rert prior to continuing.

u 1NV-lA(NC L/D bolToReqen [-lxi NV-2A (NC liD ol To Regen Hxj. CAUTION A Pu 1115 urge wil. occur when ch.rging flow is rdsed in next step. I etdowri should he established withruit delay to limb the amount of insurga

d. Eslablsh cooling to cgcncratic Hxby d. jf clarng flow to Rogencrctiic Hx performing t[e fcIlov:ing concurrently: canrot be estalisled. ThEN GO TO Step49.

c Etablislu 4 IeisL 55 3PM cIiaryiii flow by THROTTLING O°EW 1 NV-238 (Charging Line FIo Cortrol) or raising FL pump speed.

9-IRO1TLE 1NV-241 (UI Seal Vter lnj Flow Control) tD establish approximately 8 6PM seal injection flow to aach NC pump.

e. OPEN Ietdon line isolation valves as e. GO TO Step 49.

fol Iow:

1) OPEN 1kV-VS (Letdon Cont Outsice Isol).

2) OPEN 1kV-iA (NC L/t lsol To Ren Hx).

3) OPFN 1 kV-A (NC I ir IsnI To Regen Hx).

_4) OPEN ikV-35A(Vcriable LD Orifice Outlel CaM Isol).

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 86 APEO27 AA2.1O Pressurizer Pressure Control System (PZR PCS) Malfunction

t.bility to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions: (CFR: 43.5/45.13) PZR heater energizedlde-energized condition Given the following conditions on Unit 1:

1A, lB 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 group 1C is available Which ONE (1) of the following describes the required actions per Tech Spec 3.4.9, (Pressurizer)?

A. Restore PZR heater group 1A ONLY to operable status. B. Restore PZR heater group 1AAND lB ONLY to operable status. C. Restore PZR heater group 1AAND 1D ONLY to operable status. D. Restore PZR heater group 1A AND lB OR IA AND ID to operable status. Tuesday, July 13, 2010 Page 250 of 294

_____ FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 86 2586 Jeneral Discussion The 1A, lB and 1D Pressurizer heater groups have been de-energized. The remaining groups of heaters is capable of maintaining 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 IA and lB must be returned to operable status. Group 1D is not required forTS operability. ____________________ Answer A 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 and concludes that two groups of heaters have to be operable regardless of which groups. 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 B Discussion CORRECT: See explanation above. 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. If the applicant does not recall which groups are powered by emergency power supplies it is reasonable for them to conclude that the required groups are lA and lD. 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. LAUSIBLE: 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. If the applicant concludes that 1D heaters can be supplied from an emergency power supply it is reasonable for the applicant to conclude that restoring either lA and lB OR 1A and 1D will meet TS operability. Basis for meeting the KA Strict knowledge of Pressurizer heater energized/dc-energized conditions is RO level knowledge. However, given a condition were a group of Pressurizer heaters is dc-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 fhis 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 IA and IB). 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 I 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 hour Tech Specs. There are no actions in TS 3.4.9 which are required to be completed in 1 hour or less.

2) It can NOT be answered solely by knowing the LCO/TRM information listed above-the-line. This question requires the applicant to recall information from the Basis 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 Tuesday, July 13, 2010                                                                                           Page 251 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 86 2586 Development References Student References Provided arning Objectives:

1) PS-NC#24

References:

1) Tech Spec 3.4.9, Pressurizer

2) Tech Spec 3.4.9 Basis APEO27 AA2.10 Pressurizer Pressure Control System (PZR PCS) Malfunction

Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions: (CFR: 43.5 /45.13) PZR heater energized/dc-energized condition 401 -9 Comments: - RemarkslStatus 40 1-9 Comments: Change Stem 1st bullet to 1A and lB Change Choices B and D from 1D to 113 Change correct answer to D Reason: Q becomes more discriminatory. Resolution / Comments: Revised question as recommended by Lead Examiner. However, by changing question as requested, believe that answers C and D tcorrect._See attached document for_proposed fix. Tuesday, July 13, 2010 Page 252 of 294

Question 86

References:

From Tech Spec 3.4.9: Pressurizer 3.4.9 34 REACTOR COOLANT SYSTEM (RCS) 3.4.9 Pressurizer LCO 3.4.9 The pressurizer shall be OPERABLE with:

a. Pressurizer water level < 92% (1600 ftj: and

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 level A. 1 Be in MODE 3 with reactor 6 hours not within limit, trip breakers open. AND A.2 Be in ItIODE 4. 12 hours B. One required group of B.i Restore required group of 72 hours pressurizer heaters pressurizer heaters to inoperable. OPERABLE status. C. Required Action and C.1 Be in Lv1ODE 3. 6 hours associated Completion Time of Condition B not AND met. C.2 Be in MODE 4. 12 hours McGuire Units 1 and2 3.4.9-1 Amendment Nos. 184.166

From Tech Spec 3.4.9 Basis: Pressurizer B 3.4.9 BASES APPLICABLE In MODES I, 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 existence 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 of 10 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 otfsite power, as indicated in NUREG-0737 (Ref. 3), is the reason for providing an LCO. LCO The LCO requirement for the pressurizer to be OPERABLE with a water volume 1600 cubic feet, wtich is equivalent to g2%, ensures that a steam bubble exists. Limiting the [CO 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 150 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. APPLICABILITY 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 Nc. 0

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 87 EPEO38 2.4.11 Steam Generator Tube Rupture (SGTR)

PE038 GENERIC +/-now1edge of abnormal condition procedures. (CFR: 41.10/43.5/45.13) 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 (SIC Tube Leakage) has been implemented
Charging flow is 240 GPM The maximum charging flow limit specified byAP-lO is based on (1)

The basis for performing a rapid cooldown to a selected target temperature is (2) Which ONE (1) of the following completes the statements above? A. 1. preventing NV pump runout

2. to ensure that there is sufficient NC System subcooling following depressurization B. 1. preventing Regen Hx tube vibration

2. to ensure that there is sufficient NC System subcooling following depressurization C. 1. preventing NV pump runout

2. to ensure NC system temperature is below the saturation temperature for the ruptured SG PORV lift pressure D. 1. preventing Regen Hx tube vibration

2. to ensure NC system temperature is below the saturation temperature for the ruptured SG PORV lift pressure Tuesday, July 13, 2010 Page 253 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 87 2587 Jeneral 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 _______________________________ INCORRECT: See explanation above. PLAUSIBLE: Part 1 is plausible because NV flow is above the maximum charging flow allowed by AP-lO. Therefore, it is reasonable for the applicant to conclude that NV pump runout is a concern. Part 2 is correct. Answer B Discussion CORRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. LAUSIBLE: Part 1 is plausible because NV flow is above the maximum charging flow allowed by AP-lO. Therefore, it is reasonable for the plicant to conclude that NV pump runout is a concern. 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 possibility of lifting the SG PORV on the ruptured SG. For example, during the isolation of the ruptured SG, 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 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 possibility of lifting the SG PORV on the ruptured SG. For example, during the isolation of the ruptured SG, 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. Basis for meeting the KA The KIA 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 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): The question can NOT be answered by knowing systems knowledge. This is knowledge of detailed procedure content.

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

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

4) 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-lO (specifically what parameter is used to Tuesday, July 13, 2010 Page 254 of 294

_________ FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 87 2587 determine the target temperature for the rapid cooldown) and the basis (from the AP-lO Background Document) for performing the cooldown. ierefore, this is SRO level knowledge. Job Level Cognitive Level QuestionType Question Source SRO 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)

EPEO38 2.4.11 Steam Generator Tube Rupture (SGTR)

EPEO3 8 GENERIC Knowledge of abnormal condition procedures. (CFR: 41.10 / 43.5 /45.13) 401-9 Comments: RemarksiStatus sed revision for 2010 NRC Q87. Revision approved RFA 07/08/10. Tuesday, July 13, 2010 Page 255 of 294

Question 87

References:

From AP-lO Background Document Basis for Step 1: AF1I and 2;N5{001C10 (MC Leakage Within The Capacity Of Doth IIV Pumps) STEP DESCRIPTION FOR Case I. Steam Cererator rube Leakage CASE I STEP 1: PURPOSE: Restore NC System inventory as required and provide crteria to manually initiate SI. DISCUSS ION: When the operator enters this guideline, he des not necessarily know the size of the tube leak mat is in progress nor can lie expect tie size ot tne leaK to remain cons:ant. LCK sizes trot ili resL It in ent into this guiceline wll range from a few gal ons pe day to in excess of 100 gallons per minUte. The initial plant conditions will rage from full power operation to a plant shuldown. In all cases it is irriportant that the operator maintain control of pressurizer level as an indicatioa of reactor coolant system inventary and the capabi ity of the charging system to makeup for the steam generator tube leak. If is prelerable lo List this AP rather than SI if charging can keep up v.itt the leak. The analysis br AP/lO shows that the integrated leakaQe t sing the strategy of this guideline would be less Ihar if SI was actuated and the EPs were entered. Therefore, charging flow should be lrIdxiIrnLeU as iliuch as possibe Lu pieverl lie loss uf Pn Ieel arid Ulie requiteirieril Lu SI. Foi lhat reason the secnnd chargrig pump shoiili be strted In maxinhi7e charging flow if physically possible without damaging other plant equipment (i.e., Regen HX). Charqinq is increased and letdown is reducec as nessary :o main:ain Pzr level. A maximum flowrate of 2:32 GPM charçing is allowable. This will ensure there is ot excessive Regen HX lube vibration. Note th s is assum ng 32 GPM gong 10 the seals, which will limit the flow lhroigh the flegen FIX to 200 CPM during transient/accident oDeratiori (PIP M-O3-O5730. Tro conlrol board gauge for C9arging FloW reads from 0 -200 GPW. in order to maintain charging tlow on scale, the step provides guidaice to naintain charging flow less than 200 GPM. This is well below tie max mum allowable flowrale of 232 GPM. It shou d be noted here that the maximum fbwrate allowed through the Regen FIX dLlring Normal/Start Up/Shut Down cperatiori is 155 GPM. If this AP is used during a shutdown mode after isola:ing CL.&s. initiating an Si signal is not appropriate during a S/G tube leak or rupture. The emergency procedures assume CL4s are aligned open or ensure they are open In a shutcown event, opening the CLAs would cause the Pzr to rapid;y fill and make it dificult to coitrol Pr pressure. Initiating Sf1 wil also rapidly refill the Pzr with inital conditions of low NC pressure. In a shutdown mole after CLA5 are isolated. the optimum means to stop S/S tcbe leakage is to stay in this AP. It Pr level cr NC subccoling is lost. naniall aligning S/I flow would be appropriate. This is consistent with guidance proaded in AP/34 far shutdowi LOCP. Since McGuire decided to allow use of AP/lO ii any inoce, this guidance was required. Note That in lower mcdes. normal charging is likely to be aciequale Lu rriuiriUairi Fzi level, since Die NC sysLeiri will aiieiidy be pailiaII depiesuriLed. An encinsiire was proided In adcress aligning SI fIo during shiitdnwn mndec (after Cl As are 1 Page 3 of 37 Rev 8

From AP-lO Background Document Basis for Step 25: APiI and 21A15500)O1O (NC teakage Within The Capacity Of Both NV Pumps) CASE) STEP 25: PURPOSE: Detemline the target temperature that v.111 estabhsh sufficient subcooling in the NC so that the primary system will remain subcooled after pressure is decreased to stop primary-to-secondary leakage DISCUSSION: 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 cooldowri 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 EP5, the target temperature should be based on the core exit TC5. The 2OCF 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. SO pressure ranges were specified as human factors enhancement. The table has a low end target temperature related 5/0 pressure LESS THAN 300 PSIG. This makes it clear what to do if affected S/G pressure is less than this value. This was done to ensure the step is clear for lower temperatures in mode 3 and in mode 4. Note that if plant is already in mode 4. further cooldown will not be performed in the body of the AP. NC depressurization will be performed to minimize primary to secondary DIP. Final cooldown and depressurization v.111 then be performed in the post-SGTL cooldown enclosure. These enclosures address other shutdown requirements such as selecting LTOP on Pzr PORVs and aligning ND to RHR. As previously demonstrated, the pressure of the intact steam generators must be maintained less than the pressure of the affected steam generators in order to maintain NC subcooling. Since flow from the affected steam generator should be isolated! this pressure differential is established by dumping steam only from the intact steam generators in subsequent steps. It is not intended for the operator to reevaluate the required (target) core exit temperature or precisely interpolate between values listed in the table. When the required core exit temperature is reached. the intact SIG pressure should be controlled to maintain that temperature. Dont reevaluate target temperature if affected SIG pressure decays. CASE t STEP 28, 27. 28. & 29: Page22ofE7 Rev8

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 88 APEO58 AA2.02 Loss of DC Power

bility to determine and interpret the following as they apply to the Loss of DC Power: (CFR: 43.5/45.13)

,25V dc bus voltage, low/critical low, alarm Given the following conditions on Unit 1:

The unit was operating at 100% RTP when a total loss of onsite and offsite 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. 125 volts B. 1. l05vofts

2. 125 volts C. 1. ilOvolts

2. ilOvolts D. 1. lO5vofts

2. IlOvolts Tuesday, July 13, 2010 Page 256 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 88 L B jeneral 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 INCORRECT: 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 CORRECT: See explanation above. Answer C Discussion INCORRECT: 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 I is correct. art 2 is plausible because if the battery discharged to less than 110 volts surveillance SR 3.8.6.2 must be performed to verifi that battery cell paramters are within limits. Basis for meeting the KA The KA is matched because the applicant must be familiar with the minimum (low/critical low) voltage at which the vital battery must be separated 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 question can NOT be answered by knowing immediate Operator actions. There are no immediate actions associated with AP-15.

3) The question 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-15. 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: It can NOT be answered solely by knowing < 1 hour Tech Specs 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 knowledge contained in the Tech Spec Basis (specifically the minimum voltage limit for battery operability) to answer the question correctly.

Tuesday, July 13, 2010 Page 257 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION__88 L ,[veii Cognitive Level r QuestionType - Question Source SRO Memory BANK MNS Exam Bank Question API5NO1 iopnent References Student References Provided Learning Objectives: 1)AP15003

References:

I) Tech Spec 3.8.4 Basis (DC Sources Operating)

P-l5 (Loss of Vital or Aux Control Power) APEO58 AA2.02 Loss of DC Power

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 Comments: RemarkslStatus E 401-9 Comments: No comment. Resolution I Comments: N/A Tuesday, July 13, 2010 Page 258 of 294

Question 88

References:

From Tech Spec 3.8.4 Basis: D aurceeOpera1Irg B 3.9.4 BASES DCKC-RQUND canbnue: Eaofl cattery EVG& EVOb, EVOC, EVODi has adequate sirage capacty to carry The required duty cce ror one hour alter the loss or the balt@fy charqer outtiuL [n adaiion. the bafley is eapanie 01 suppqng power ft lt,e operatlor ol arliclpated rnomen tary oads duhrg The one hour pr ec. Each 125 VDt battery is separately housed fri a enHla1ed room apart torn Its oh arger xd dishibut on ceriers. Each channel is iocatel in an area separated phys.caii and eieciroalyltorr the oTher cnarnei to ersure that as ngieiaiure in one subsystem does nd cause a raliure In a redurdail e.ubsysleni. There is no sharlrg betseer redundar.t Class 1E subsysterris suct as nattenes, battery chargers, ordistrbulion careit Th tisres irr vhanri f D are size w prauce reçuir capacty at 80% oinanepiate ratng, ccnespondng to warranted capacty at ena 01 iiie oyaies aid tue 1 r]% desgn deland. aitery ae baseli or 125% or required capacty ard, alter seieciion or an .aiaiiaale cornn&cai flry. rsu[ts in a ati&iy taaacity in &XC4.b OV 15E% at required capaciy. The hidivttaiai ceii voltage uiit Is 2.13 V per oeii. The rlr:rm naervrrnlra uotellrlt s rea1er than reuait9 1261.1

                              .iitie on roatosiarge asdiscussel n the LIFSAR, Cliap:er S (ReL :*

Ths nterrotsrg arg .ea- storags bat1&resre defMe ir. IEEE--LBS aflst 5]. Each chanrei of DC has anp.e poaer output cap cty ror the steady state operailon or ocinected lcaads required during norniai op.eraUon, wh.e at the same time nairtanirg ts battery ba9k ltii y chargers. Each batlery chareraso hassisllcieritcap.acty to restore the battery hon the desIgn mm turn harpie Q iteiu.y charged ste within e hcur athle sia.apiyitiq romiai steady state loads discussed in the UFSAR Chanter 8 Ret APPL1DAELE The iwai esnu lions or Desigi Basis Accident (DaA ard transient SAFETY ANALYSES araiyses i Ihe uFSAR, Cihapler 6 i:Ret 5j, and ir the UFSkR, Chazter 15 (R8-t 7], nnrn ihat Enhiesred Oiety Fealure EOF) sysleTs are ORE RAELE The OPERABiLiTY otihe DC sources is consistent wth the tnitai assurnpflons or ihe accident anayses ard e based upon rneeiing the n basis ot me unit This i nciudes rairia nirg The DC sources o PERABLE du nrg acedent cordit ens In the eve!lt at: Mosuhre Uiits 1 ard 2 0 aMA-2 eviion ND. iCJ

From AP-15 (Loss of Vital orAux Control Power): MNS LOSS OF VITAL OR AUX CONTROL POWER PAGE NO. AFI1IA/SSOQl5 of 268 Cnclosure I Pge 14 of 17

tNIT 1 Response To Degraded DC Bus Voltage e. 2DI1on.:Exrz:TE:I flZCflE rz;rc*:i-z :o: cann;z: NOTE l arI distrbution center vdtaçe goes down to 105 107 /ots, the associated

battery ha met is cLay cycle requirement. Further cepletion cf the battery may rsLlIt in hr4temy rhnnge Cpning the scviited hM1er hrpnker when this voltage is reached vAIl complelely deenergize the associated distribution certer

17. IF AT ANY TIME dispatched operator notifies Control Room that distribution center voltage reaches limrt in table bdOW, ThEM GO TO indicated step to remote battery from service:

DISTRIBUTIGN VOLTAPE STEP ut11nR L_N.. -. Step IS fl Step 23 z7:L. 13; V*zlt; 21 E1D 135 V SLt? 22 EVOC I:Lt V:its Step 23 EVDD LOS V:its Step 24

18. Do not continue unlcss dlr,ctad cy Step 17.

MNS LOSS OF VITAL OR AUX CONTROL POWER PAGE NO. APIIIA/5500115 65 of 268 Enclosure 1 -Pao 16 of 17 UNIT 1 Response To Degraded DC Bus Voltage Rev. 20 ATION/EXEDTD RE5PONE RESPONSE NOT OSTAINED

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. IFAT,ANYTIME control power is needed to operate A Train breakers, ThIEN, 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 TO AP2/A1551J0,15 (Loss of Vital or Aux Control Power) as time allows.

2) Trip Unit 2 reactor

3) GO TO EP2A/5000iE-O (Reactor Trip or Safety Injection).

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

e. Trip Unit 1 reactor.

f. T EPI1!A50001E-O (Reactor Trip or Safety Injection).

22. JF AT ANYTIME EVDB Distribution Center reaches 105 volts, THEN evaluate performing the following:

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

b. Notify Unit 2 to GOTQAP!2JA5500/15 (Loss of Vital or Aux Control Power).

c. RETURN TO 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:

I 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 1 EVDA is designed to sustain loads and what action will protect the DC bus loads?

A. After 1 hour, the vital battery bus breaker will open automatically when bus voltage falls to 105 volts. B. After 1 hour, the vital battery breaker must be manually opened when bus voltage falls to 105 volts. C. After 4 hours, the vital battery breaker will open automatically when bus voltage falls to 107 volts. D. After 4 hours, 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

hour 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 hour and there is no automatic trip associated with low voltage Plausible: the 4 hour 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 hours Plausible: partially correct DC bus protection is achieved by

manually opening the breaker voltage limit is 107 volts.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 89 APEO62 2.4.47 Loss of Nuclear Service Water

PE062 GENERIC Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. (CFR: 41.10

/43.5 /45.12)

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 L 4:00 5:00 6:00 7:00 8:00 9:00 10:00 11:00 Based on these conditions, the SNSWP level becomes initially 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 Tuesday, July 13, 2010 Page 259 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 89 General Discussion In the scenario the applicant is presented with a set of conditions where a loss LU 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 SXSWP 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 I 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 jterpret this to imply that the safe shutdown loads supplied would be required to be available for the same time frame. Answer B Discussion iORRECT: See explanation above PLAUSIBLE: 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 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 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 nterpret 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 iis 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 -. question meets the following criteria for an SRO only question as described in the Clarification Guidance for SRO-only Questions Revi 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 hour Tech Specs. There are no actions that are 1 hour 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 Tuesday, July 13, 2010 Page 260 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION_89 Development References ntReferencesProvided 3.7.8 Basis AP-20 Enclosure 1 APEO62 2.4.47 Loss of Nuclear Service Water

APEO62 GENERIC Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. (CFR: 41.10 /43.5/45.12) 401-9 Comments: RemarkslStatus 401-9 Comments: Add the word initially before the word TNOPERA.BLE. Otherwise, if the first part of B and D were correct, A and C would be correct too. The Q is a U because there are potentially 2 correct answers if the correct time was 1040 for the first part of the question. Resolution I Comments: Added the word initially per Lead Examiners comment. See attached file. Tuesday, July 13, 2010 Page 261 of 294

Question 89

References:

From TS 3.7.8 Basis: 3N3VP E 3.7.B BASES APPLICAELE SAFETY AN1kLYSES hontinued) decay heat, anc worst case sinqie actvefalure (e.., single failure cf a manmade structure). The SNSWP is designed in atcordance wib Regulatrv Guide 1.27 Ref. 2), thch reures a 30 day supply of coolwiq wate in the SNSVcP. The SNSVJPsctslies Critehon S of 10 CFR 50.33 (Ref. 3). LCO The SNSWP is equirsd to be ODA9LE and is cons idcred OPER.BLE Th is the hasb if it conbins a siffiDient volume of water at or below the maximum containing the correc: ternperatwe that voud allow the NEW. S to Dperate for at least 24] deys answer .,hich is 30 days. F following the desigi Lasis LODA without the loss of net psiive siction bead (NPSI l. and wi:hcutexr the rnaxmun &-n 1ernprature _JI. L r_.n --

._

APPLICAFILIIY In MODES 1,2,3, arid, the SNSWP is requVed to suprnr*. OPERABILITY of the eqiiprnerit serviced by the SNSWF Tht ecdoz coatass +/-e be OPERABLE in these MODES. livelwbehr ?395 ft. The take forte ]aveforthe In MODEL or 6, the recpirements of teSNSWP are del runedtt2nsen 722 nck pxande ?1aubht, fcr 5 BtEffi5 it suorts rr theappltartElectünztke ka adU4ILc. ACTIONS A. I If the St46P is inperciblc the uni: must be pad in a MODE in which the LCO daes not apply. To acheve this statis, the unit must e placed in at least MODE 2 wthin 6 hours and in MODE 5 tFün 36 host. The allowed Coiieton Times are reasonable, based on operating eçerience, t-) reach the requ red wit coidio,s fran ulI pcwer in iditiuis iii an oideily Irk.1i ii er Lid iU uut cliulleruir y uiiil sysleiiii SUR/ELLANCE SR 3.78.1 RhUIJIRtMLN IS This SF: verifes that adeqiate long te-m (30 thyi cool ng can be i,.,ni .,i. .-..., .,.,. ureta

From AP-20 Enclosure I MNS LOSS OF RN PAGE NC). AP/ l/A15500i2C1 Enclosure 1 Page 1 of 4

UNIT 1 Aligning B Train RN to Pond e. NOTh Shared RN valves can be operated from either units control switch. Align B Train RN to SNSWP as follows:

a. Notify Unit 2 Operator that B Train RN will be aligned to the SNSWP.

b. Check ORN-95 (train lB & 25 SNSWP b. GQ TQ Case II (Loss of Low Level Supply) OPEN.

                        -                                          or RC Supply Crossover).

c Open ORN-1525 (Train 15 & 25 Disch c. Perform the following to minimize pond to SNSWP). depletion:

1) Close the following valves:

                                                                   . ORN-ii5 (Train 1B&2BLLI Supply)

ORN-i OAC (train I B & 25 LLI Supply).

2) Dispatch operator to open ORN- 1525 (Train 15 & 25 Disch to SNSWP) (aux bldg pipe chase, 7i6÷Z EE-64, near Unit 2 containment surnp lines 15 from cad door)

_3) WHEN QRr4-1526 ([[rain 15825 Disch to SNSWP) is open, THEN perform Steps I d through 1. h

4) Monitor SNSWP level as required to pre.ent reducing SNSWP below Tech Spec level.

(RNO continued on next page.:

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 90 L 2590 APEOO3 2.2.40 Dropped Control Rod

. PE003 GENERIC bility to apply Technical Specifications for a system. (CFR: 41 10 /43.2 / 43.5 /45.3) 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 hours.

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 Tuesday, July 13, 2010 Page 262 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 90 259O

  ,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 hour OR SDM must be verified with limits AND power reduced to less than 75% RTP within 2 hours.

In addition surviellances for Enthalpy Rise and Heat Flux Hot Channel factors must be performed within 72 hours. 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 hours 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 performed. 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 hours. Answer B Discussion ORRECT: 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 ORRECT: See explanation above. Answer DDiscussion 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 hours 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 hour 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 hour. 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 I-li Cog Basis for SRO only Part I 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)(2) (Tech Specs): I) This question can NOT be answered by knowing less than 1 hour 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 ofTS 3.1.4.

3) This question can NOT be answered by knowing the TS Safety Limits or their bases. This question relates toTS 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 luction 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 I dated 03/11/2010 for screening questions linked to IOCFR55.43(b)(5) (Assessment and selection of procedures): Tuesday, July 13, 2010 Page 263 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 90 2590 C

1) The question can NOT be answered solely by knowing systems knowledge. The requirements for retreiving a dropped rod are not discusse limits and precautions or in the systems lesson plan. Therefore, this is NOT systems knowledge.

The question can NOT be answered by knowing immediate operator actions. The action to reduce power to less than 50% RTP to recover the uropped 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 L I I evelopment References udent_References Provii Learning Objective: 1)AP-14#4

References:

1) Lesson Plan OP-MC-IC-IRE APEOO3 2.2.40 Dropped Control Rod

APEOO3 GENERIC Ability to apply Technical Specifications for a system. (CFR: 41.10/43.2/43.5/45.3) 401-9 Comments: RemarkslStatus

                                                                                                                                                   -1 40 1-9 Comments:

Due to the fact the Q is written in two parts renders distractor NP. Change Dl to something other then 75. Resolution I Comments: Changed distracter D to 50% RTP as this was the only plausible answer remaining. See attached file. If this is acceptable need to revise the distracter analysis for answer D. Tuesday, July 13, 2010 Page 264 of 294

Question 90

References:

From AP-14: MNS ROD CONTROL MALFUNCTION PAGE NO. API I iNS 50 14 of 37 Endosure 1 Page 5 of 23

UNIT 1 Response To Dropped or Misaligned Red Rev. 13

                     -_:::_:::D:::::::DDD:::

12. Reduce reactot power below 50% prior to rod realignment as follows:

a. Checkonlvone rod-MISALIGNED. a. QIQStep 12.c.

b. Ensure reactor power is loss than 75%

within 2 hours of rod nisaliqnrnent to comply with Tech Spec 3.1.4.

c. Reduce load as directed in subsequent steps until reactor power is less than 50% to comp with Reactor Engineering requirements.

d. Observe the following limitations during power reduction:

I) Do not move rods until IAE determines rod movement is available.

2) Borate as required during power reduction to maintain T-Avg at T-Ref.

3 Monitor AFD during load reduction.

4) IFATANYTIME AFD reaches Tech Spec limit AND reactor po.er is greater than 50%. THEN, perform the following:

a) Trip Reactor. b) GO TO EP/1LAJS000/E-0 (Reactor Trip or Safety lrection).

e. Reduce reactor power to less than 50%

PER one of the following procedures: OP/1LAJG 1001003 (Controlling Procedure For Unit Operation). Enclosure 4.2 (Power Reduction) OR AP/ IIA,5500104 (Rapid Downpower).

From TS 3.1.4 Basis: Rod Group Alignment Limits B 3.1.4 BASES ACTIONS (continued) 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 power reduction, since control bank D must be moved fully in and control bank C must be moved in to approximately 100 to 115 steps. Power operation may continue with one RCCA trippable but misaligned, provided that 5DM is verified within 1 hour The Completion Time of 1 hour represents the time necessary for determining the actual unit SDM and, if necessary, aligning and starting the necessary systems and components to initiate boration. 6.2.2, 6.2.3, 6.2.4, 6.2.5. and 6.2.6 For continued operation with a misaligned rod. RTP must be reduced, 5DM must periodically be verified within limits, hot channel factors Fc(X.Y,Z) and F(X,Y) must be verified within limits, and the safety analyses must be re-evaluated to confirm continued operation is pemiissible. Reduction of power to 75% RTP ensures that local LHR increases due to a misaligned RCCA will not cause the core design criteria to be exceeded (Ref 7). The Completion Time of 2 hours gives the operator sufficient time to accomplish an orderly power reduction without challenging the Reactor Protection System. When a rod is known to be misaligned, there is a potential to impact the 5DM. Since the core conditions can change with time, periodic verification of 5DM is required. A Frequency of 12 hours is sufficient to ensure this requirement continues to be met. Verifying that F.CX.Y,Zj and Fs.(X.Y) are within the required limits 3 ensures that current operation at 75% RTP with a rod misaligned is not resulting in power distributions that may invalidate safety analysis assumptions at full power. The Completion Time of 72 hours allows sufficient time to obtain flux maps of the core power distribution using the incore flux mapping system and to calculate Fo(X.Y.Z) and Once current conditions have been verified acceptable. time is available to perform evaluations of accident analysis to detemine that core limits will not be exceeded during a Desn Basis Event for the duration of Mcsuire Units 1 and 2 B 3.1.4-6 Revision No. 0

From TS 3.1.4: RDd GroLp Aignnient Limits 3.1-4 ACTIONS (continued) CONDITION REQUIRED ACTION CDMPLEflON4 TIME

8. One rod not withn 8.1 Restore rod to within 1 OLl1 alignrnert limits, a ignment limits.

OR 8.2.1.1 Verify 5DM i3 within the 1 iour limit specified in the COLR. OR 8.2.1.2 Initia:e boration to restore 1 rour 5DM to cyithin linit. AND 8.2.2 Reduce THERMAL 2 lours POWERto< 75% RTP. AND 8.2.3 Vei1y 5DM is widriri tIre Oirce per Iinrit specified in the DOLR. 12 hours AND 8.2. Perfcrm SR 3.2.1.1. 72 hours AND 8.2.5 Perlurii SR 3.2.2.1. 72 Iioui AND 8.2.6 Re-eqaluate safety 5 days analyses and confirm tsuIIs r eriiiri valid lvi duration of oDeratior under these coiditions. (continued) McGuire Units 1 and 2 3.14-2 Arnendrrrert Ncs. 184/166

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 91 259l APEO69 AA2.02 Loss of Containment Integrity

bility to determine and interpret the following as they apply to the Loss of Containment Integrity: (CFR: 43.5 / 45.13) verification of automatic and manual means of restoring integrity Given the following conditions on Unit 1:

The unit is in MODE 5 following a refueling outage
PT111A142001002 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. Tuesday, July 13, 2010 Page 265 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 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, [nd 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

NC0T 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 -__________________________________________ INCORRECT: 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 this 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 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 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-l9.

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-19, Loss of ND orND System Leakage -

E069 AA2.02 Loss of Containment Integrity

Ability to determine and interpret the following as they apply to the Loss of Containment Integrity: (CFR: 43.5 / 45.13) Verification of automatic and manual means of restoring integrity Tuesday, July 13, 2010 Page 266 of 294

_________ FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 91 2591

ó9 Comments:          -

I Remarks/Status 401-9 Comments: Is it ever possible that the WCC SRO is ever the containment closure coordinator? If so B could be a potential correct answer. FAC please confirm one way or the other. This Q is E until confirmed. Resolution / Comments: The WCC SRO is never used to perform the function of the Containment Closure Coordinator. The Containment Closure Coordinator position is filled by individuals separate from the

                                   çposition.

Tuesday, July 13, 2010 Page 267 of 294

Question 91

References:

From AP-19: MNS LOSS OF ND OR ND SYSTEM LEAKAGE PAGE NO. AP?1:Ns5D19 B of 217 UNIT] Rev 22

zz:::::::::z::::: :DDz::Ecmz::ccc:

5. Evaluate isolating Dont4nnlcnt as fnl1na a Checc both ND pumps OFF. - a. Perform tie folloviinq:

1) leak ze greer than 10 GPW, IttFgGO l(JSlepb.i

2) i.E leak caused trip 2 alan, zn arty Ccntcinrne,t or lJnt veni EMF.

J THEIGOTOStcp&i

3) EAIAIIYJ1MF hnth Nfl pumps off, ]IIEM perform Step &b throuqh S.f.

4) i.E MMYJIML leak size greater lhm 0 GFM OR leak caused trip 2 alarm or anr Dmtainnieit or lint vent LW, perform Steps Cb through S.f

5) fiQI2StepG.

Ii Annoince the fdLo;irq on Pale:

1) Des en pticn ot event.

2) All persomel evacuate Uiit I cont3i nrner,t.

c. Actuste contsinrrient evacuatien alem.

d RFFFR TO RPI4EAR?flCllIl II Conducting a Ste Asse,tly, Sie Evacuator. ci Contaiinent Evacuator while eortinJlrg vith this procedure.

e. Cliecs P17IIA42001002C _e. GOTOStep6.

Containment Cbsure) IN LrrLCT.

f. Notif Czntainmeni Closure oordinabr to nitate con:ainnl?ni closure.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 92 2592 APEO76 24.11 High Reactor Coolant Activity

( PE076 GENERIC

  .now1edge of abnormal condition procedures. (CFR: 41.10 I 43.5 I 45.13)

Given the following conditions on Unit 1:

The unit is at 100% RTP
AP-1 8 (High Coolant Activity) has been entered due to 1 EM F-i 8 (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. failed fuel

2. increase letdown flow C. 1. crud burst

2. place the Cation Bed demineralizer in service D. 1. crud burst

2. increase letdown flow Tuesday, July 13, 2010 Page 268 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 92 2592 eneral Discussion From the basis document for AP-l8: 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 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 B 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 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- 18 which are specific to a crud burst or failed fuel event. Answer D Discussion CORRECT: See explanation above. Basis for meeting the KA he KA is matched by Part 2 of the question in that the actions listed are from AP-l8 and are different for failed fuel as opposed to a crud burst. rt 1 of the question asks for information from the Background Document for AP-l8. 3asis 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 10CFR55.43(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-l8 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 -- arning Objectives:

 ,AP18003

References:

1) AP-18

2) AP- 18 Background Document Tuesday, July 13, 2010 Page 269 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 92 2592. kPE076 2.4.11 High Reactor Coolant Activity

E076 GENERIC nowledge of abnormal condition procedures. (CFR: 41.10 / 43.5 I 45.13) 401-9 Comments: RemarkslStatus 401-9 Comments: Failed fuel is NP for Cobolt and Manganese. Either rewrite the stem or change distractors A and B. The Q is U because of 2 NP distractors. Resolution I Comments: Believe that there is plausibility since two validators picked failed fuel as the correct answer. Went through AP- 18 and AP- 18 Background document again to determine if there is another possible direction for the first part of this question. There is very little content in AP-18 and very few options for replacing the first part without testing something that is minutia with little operational validity. Tuesday, July 13, 2010 Page 270 of 294

Question 92

References:

From AP-18 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 hours 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 hours (at 120 gpm letdown flow, 50% of the crud is removed every 12 hours).

REFERENCES:

Primary Chemistry Lesson Plan OP-MC-CH-PC

From AP-18: MNS HIGH ACTIVITY IN REACTOR COOLANT PAGE NO. AR LA15500118 2 of 4 ITNIT 1 Rev. 3 DTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED B. Symptoms

1EMF-48 REACTOR COOLANT HI RAD alarm
IEMF-iS REACTOR COOLANT FILTER 1A alarm
1EMF-19 REACTOR COOLANT FILTER 1B alarm
Chemistry sample results indicate an unexpected increase in NC System activity.

C. Ooerator 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 DEMI N position.

Temp Cntrl) ALIGNED TO DEMIN.

3. Determine cause of high activity as follows:

a. Request Chemistry to check ciecontanlinetion factor of mixed bed dern nerd izer.

b. Notify Chemistry to perform all NC System isotopic analysis to determine it high activity is from a crud burst or tailed fuel.

4. IF AT ANYTIME it is determined that high activity is from crud burst, THEN raise letdown flow to 120 GPM PER 0P111A16200!O01 A (Chemical and Volume Control System Letdown),

Enclosure 4.5 (Establishing Maximum Normal Letdown).

MNS HIGH ACTIVITY IN REACTOR COOLANT PAGE NO. AP/1/A/5500/18 3 of 4 Rev. 3 UNIT 1 AcTIcN/EX?EcDD EsONsE P.ESPONSE NOT OBTAINED

5. IF AT ANY]WE it is determined that high activity is from failed fuel, THEN perform the following:

a. Ensure mixed bed dernineralizer 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. fflA[ANYflM Chemistry requests cation bed demineralizer be placed in service, THEN place in service PER OP/I !A/6200/OO 1 D (Chemical and Volume Control System Demineralizers). Enclosure 4.3 (Removing/Returning the Cation Bed Demineralizer from/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. REFER [ Tech Spec 3.4.16 (RCS Specific Activity).

8. WHEN station management determines Control Room pressurization no longer required, THEN secure PER 0P101A16450101 1 (Control Area Ventilation)Chllled Water System),

Enclosure 4A (Control Room Atmosphere Pressurization During Abnormal Conditions). END

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 93

.WE03 2.4.46 LOCA Cooldown and Depressurization
                   -

( 7E03 GENERIC bi1ity to verifi that the alarms are consistent with the plant conditions. (CFR: 41.10 / 43.5 / 45.3 / 45.12) Given the following conditions on Unit 1:

A Reactor Trip and Safety Injection have occurred due to a Small-Break LOCA inside Containment
Containment pressure peaked at 2.5 PSIG
ES-1.2 (Post LOCA Cooldown and Depressurization) has been implemented
Both ND pumps are running
NC system pressure is 250 PSIG and decreasing slowly The FIRST FWST level and Containment Sump conditions that require stopping both ND pumps prior to swapping to the containment sump are FWST level (1) AND both CONT SUMP LEVEL GREATER THAN 2.5 FT alarms are (2)

Which ONE (1) of the following completes the statement above? A. 1. 200inches

2. DARK B. 1. 260 inches

2. DARK C. 1. 200 inches

2. LIT D. 1. 260inches

2. LIT Tuesday, July 13, 2010 Page 271 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 93

.eneral Discussion In accordance with ES-1.2, if FWST level decrease to less than 250 inches AND both CONT SUMP LEVEL GREATER TI-lAN 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 INCORRECT: See explanation above. 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 t understand that there has to be sufficient inventory in the Containment Sump prior to swapover to prevent vortexing of the ND pumps, the ond 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 Hi 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 IOCFR55.43(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-l.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 SRO Cognitive Level Comprehension J QuestionType NEW Question Source eveIopment References Student References Provided Learning Objectives: 1)EPEIOO4 Tuesday, July 13, 2010 Page 272 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 93 2593

References:

E 1 Background Document WEO3 2.4.46 LOCA Cooldown and Depressurization

WEO3 GENERIC Ability to verify that the alarms are consistent with the plant conditions. (CFR: 41.10 / 43.5 / 45.3/45.12) 401-9 Comments: RemarkslStatus 401-9 Comments: I fail to see the connection between this Q and LOCA cooldown and depressurization. The stem bullets involve it by referencing ES-1.2, but the Q stem and choices do not. This Q is a U until FAC justifies. Resolution / Comments: In ES-1.2, the check for FWST level less than 250 inches with the Containment Sump Level Alarms DARK is a continuous action step to ensure that there is sufficent level in the sump to prevent vortexing at the suction of the ND pumps when the auto swapover level in the FWST is reached. This is an important step during Post LOCA Cooldown and Depressurization. Tuesday, July 13, 2010 Page 273 of 294

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 416OVAC 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 II, 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 SIG levels: (CONTINUOUS ACTION) PURPOSE: To ensure adequate feed flow or S/G 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 SIG 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.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 94 - GEN2.I 2.1.4 GENERIC Conduct of Operations

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( onduct of Operations i<nowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, IOCFR55, etc. (CFR: 41.10 /43.2) Unit 1 is operating at 100% RTP. An active licensed STA may assume the duties of the Control Room Supervisor provided 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. 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 Tuesday, July 13, 2010 Page 274 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 94 2594 General 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 hour for the entire shift.

Answer A Discussion ____________________________________________________________ RRECT: See explanation above. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant 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: See explanation above. PLAUSIBLE: This answer is plausible if the applicant 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 D Discussion INCORRECT: See explanation above. JPLAUSIBLE: This answer is plausible if the applicant confuses the time for the CRSRO or relief SRO to return to the control room with the lowable duration of the relief by the STA. asis 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 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)(1 & 2) (Tech Specs):

1) This question can NOT be answered by knowing less than 1 hour 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 ferences:

\.. ,Technical Specification 5.1.2, amendment 213 and 194 GEN2.1 2.1.4 GENERIC Conduct of Operations

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Conduct of Operations Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, Tuesday, July 13, 2010 Page 275 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 94 2594 maintenance of active license status, 1 OCFR55. etc. (CFR: 41 10 I 43.2) J1-9 Comments: Remarks!Status 401-9 Comments: For this particular question, since the times are the same for C and D, do not make this a fill in the blank. Consider writing C and D as follows: An active licensed STA may assume the duties of the CRS provided the relief SRO is available to both return to the control room AND the periods which the STA assumes SRO duties do not exceed 10/15 minutes in duration respectively Otherwise, C and D will be ruled out due to the way they are worded. This Q is E until modified. Resolution I Comments: Reworded per Lead Examiners suggestion. See attached file for proposed revision. As a note, two validators missed this question and BOTH picked C. Tuesday, July 13, 2010 Page 276 of 294

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)] th 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 SRO to be absent from the control room in MODE 1 2, 3, or 4, 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 SRO 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 hour during any shift, and

c. the STA has a SRO license on the unit.

McGuire Units 1 and 2 5.1-1 Amendment Nos. 213/194

Question 94 Parent Question (MNS 2009 NRC Exam): 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 J 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 hour 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 bytheSTA. 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 bytheSTA. 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 hour 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.

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 95 GEN2.1 2.1.8 GENERIC Conduct of Operations

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onduct of Operations Ability to coordinate personnel activities outside the control room. (CFR: 41.10/45.5/45.12/45.13) 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 1 KF-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). Tuesday, July 13, 2010 Page 277 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 95

  ..éneraI 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 1KF-l22.

Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because this would be the correct answer if 1KF-122 could not be closed. However, the first attempt is to close 1KF-l22. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible because the Operators are directed to make upt o the spent fuel pooi in AP-40. However, the first action is to attempt to isolate the spent fuel pool from the refueling canal the preserve the water that is in the spent fuel pool. Also, makeup to the spent fuel pool 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 D Discussion CORRECT: See explanation above. Basis for meeting theKA 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 asis 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. 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-41.

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 ofAP-41 and into AP-40 (Loss of Refueling Canal Level) where they are directed to perform the appropriate actions.

Job Level Cognitive Level jQuestionType Question Source SRO Memory BANK MNS Exam Bank Question FHFCNO14 Development References udent References Provided Learning Objective: 1) ?) eferences: esson Plan OP-MC-FH-FC Section 3.2.2 AP-40 GEN2.1 2.1.8 GENERIC Conduct of Operations

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Conduct of Operations Ability to coordinate personnel activities outside the control room. (CFR: 41.10 / 45.5 / 45.12 / 45.13) Tuesday, July 13, 2010 Page 278 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 95 259 1-9 Comments: iemarksIStatus - - 40 1-9 Comments:

                                 ,No comment.

Resolution / Comments: N/A Tuesday, July 13, 2010 Page 279 of 294

Question 95

References:

From Lesson Plan OP-MC-FH-FC Section 3.22: 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 RPIOIAI5700I1 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 RPIO/A15700100, 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 0P111A16200/13. CAUTION: Makeup to the SFP 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. APi 1 IA!5500140 2 of 18 UNIT 1 Rev. 7 AION!EXPECTED RESPONSE RESPONSE MOP OBTINEO B. Symptoms

SPENT FUEL POOL LEVEL LOW computer alarm
Level in refueling cavity going down
INCORE INST ROOM SUMP HI LEVEL alarm
I EMF-16 CONTAINMENT REFUELING BRDG alarm
1 EMF-17 SPENT FUEL BUILDING BRDG alarm.

C Operator Actions

1. Announce occurrence on pace.

2.

                      -L Check FUEL MOVEMENT IN
                    -                                       Perform the following:

PROGRESS.

a. IF any radioactive component is being handled in 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 AP.

c. GOTOStep4.

MNS LOSS OF REFUELING CAVITY LEVEL PAGE NO. API1!A15500140 3 of 18 Rev. 7 uxn 1 CTIONJEXPECD ESPDNSE RESPONSE NOT OBTAINED NOTE Any availab[e core location may be used when lowering a fuel assembly during emergency conditions.

3. Contact fuel handling SRO to have fuel handling crew perform the fol lowing:

-K

a. Lower any fuel assembly in the reactor a. Release brake and hand crank hoEst building manipulator crane to fully down down.

in the core.

b. Lower any fuel assembly in the spent b. Release brake and hand crank hoist fuel nanipulator crane to fully down. down.

4

c. Lower any fuel assembly in either c. Release brake and hand crank upender upender to fully down. down.

ci. fuel transfer cart to the spent tuel ci. Release brake and hand crank transfer (Pit) side, cart to spent fuel (Pit) side. 4

e. Lower aiiy radioactive component in the e. Perform the following:

spent fuel pool or refueling cavity to fully down.

Reinstall component.

I CR

                                                          . Place component as far below the water us safely possible.

4. WHEN fuel transfer cart is in the spent fuel bldg, THEN dispatch 2 operators to CLOSE IKF-122 (Unit I Fuel Transfer Tube Isol) (spent fuel bldg, 780, PP-51, top of fuel pool at south east corner).

5. Notify Containment Closure Coordinator to initiate containment closure PER PT111N42001002 C (Containment Closure).

Question 95 Parent Question: FHFCN014 1 Pt(s) Given the following conditions:

Unit I 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 1 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: OPMC-FH-FCB TASK:

p. 59, 60 OBJECTIVE: 15C TIME:

K/A: 000036G010 (3.7/3.8) DATE: SROEXAM1992

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 96 2596 GEN2.2 2.2.40 GENERIC Equipment Control

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quipment Control Ability to apply Technical Specifications for a system. (CFR: 41.10 /43.2 /43.5 /45.3) Given the following conditions on Unit 1:

          . Unit 1 is operating at 100% RTP The following sequence of events occurs:

DATE TIME EVENT 0800 July10 1SA-49AB (Main Steam Supply from SG lB to TD CA Pump) declared INOPERABLE 0800 July 16 1SA-48ABC (Main Steam Supply from SG 1C to TD CA Pump) declared INOPERABLE 0100 July 17 1SA-49AB (Main Steam Supply from SG 1 B to TD CA Pump) returned to OPERABLE status In accordance with Tech Spec 3.7.5 (AFW System), 1SA-48ABC must be returned to OPERABLE status by or the unit must be placed in MODE 3 within 6 hours and MODE 4 within 12 hours. Which ONE (1) of the following completes the statement above? REFERENCE PROVIDED A. O800on July17 B. 0800 on July 20 C. 0800 on July 23 D. OlOOonJuly24 Tuesday, July 13, 2010 Page 280 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 96 L2596 eneraI Discussion In accordance with TS 3.7.5 (AFW System) the original T.S. entry on July 10 at 0800 for 1SA-49AB expires on July 17 at 0800. The second steam supply from ISA-48ABC becomes inoperable at 0800 on July 16. Since both steam supplies are inoperable concurrently, the rule for 10 Ldays from discovery of failure to meet the LCO applies. So, after 1 SA-49AB is returned to service on July 17 at 0100, the 10 day LCO [requirement from the time of the initial entry into the T.S. would require action be taken on July 20 at 0800. Answer A Discussion ____________________________________________ INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant does not understand the 10 day allowance in the spec and concludes that the 7 day LCO requirement still applies from date and time of the original entry into the spec. That being the case, the applicant would determine that this is the correct answer. Answer B Discussion CORRECT: See explanation above. Answer C Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant concludes that after 1SA-49A13 is returned to service, the 7 day LCO requirement applies from the time that I SA-48ABC became inoperable (i.e. the steam supply that is still inoperable). That being the case, the applicant would determine than action must be taken on July 23 at 0800. - Answer D Discussion INCORRECT: See explanation above. PLAUSIBLE: This answer is plausible if the applicant concludes that after 1SA-49AB is returned to service the LCO becomes 7 days from the time that 1SA-49AB is returned to service (1SA48ABC is still inoperable). The applicant would then conclude that action must be taken on July 24atOlOO. Basis for meeting the KA e KA is matched because the applicant, given a copy of Tech Spec 3.7.5, to apply the specification to given data and determince the correct ,CO time. Basis for Hi Cog This is a higher cognitive level question because it requires multiple mental steps to arrive at the correct answer. First the applicant must recall the rules of usage for applying the 10 day extension from T.S. Basis. The applicant must then analyze the equimpent inoperability times to determine the correct action time. Basis for SRO only 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 hour Tech Specs. The only part of T.S. 3.7.5 that requires action in less than 1 hour is if all three AFW trains are inoperable in MODES 1, 2, or 3 OR if the single AFW train required to be operable in MODE 4 is inoperable.

2) This question can NOT be answered by knowing information listed above-the-line. This question requires the applicant to apply T.S. 3.7.5 requirements that are below-the-line.

3) This question can NOT be answered by knowing the TS Safety Limits or their bases. This question requires the applicant to apply T.S. 3.7.5 and recall usage requirements from the T.S. Bases (NOT Safety Limit bases).

4) This question requires the applicant to apply T.S. 3.7.5 and recall usage requirements from the T.S. Bases. Therefore, it is SRO level knowledge.

Job Level Cognitive Level fQuestionType Question Source SRO Comprehension BANK MNS Exam Bank Question ADMTSO2O Development References Student References Provided ch Spec 3.7.5 Tech Spec 3.7.5 (AFW System)

  .ch Spec 1.3 Completion Times GEN2.2 2.2.40 GENERIC Equipment Control
                    -              -

Equipment Control Ability to apply Technical Specifications for a system. (CFR: 41.10 / 43.2 / 43.5 /45.3) Tuesday, July 13, 2010 Page 281 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 96 2596 1-9 Comments: I RemarkslStatus Proposed replacement for 2010 NRC Q-96. Replacement question approved. RFA 07/06/10 Tuesday, July 13, 2010 Page 282 of 294

Question 96 High Miss Proposed Replacement

References:

From T.S. 3.7.5: AFW Svstrn

                                                                                                  .37.5 31 PLANT SYSTEMS 37.5 Auxilii.i y Feecwiler (4.FW) SysWiii LIX) 3.7.5            Tlree AFW trains shall be OPERAbLE Only one AFvV train. whi:h helides a notor dhven pimp, is squired to be OPCRAIJLC in MDD[ 4.

APPUCABILTY MODES 1, 2, anc 3, MODE 4 then stean qeterator is relied upon for heat rernoaL ACTIONS LCD 3M.4.b is not apIicab Men ensring MODE 1. CUNUI [ION REUUlftL) PU I ON UONPLh I IUN l A. One steam supDIy to A. 1 flestore steam supply to 7 days turbine driven AFW OPERABLE status. pump inopeable. ND PDda,t ro rFt Steam ci sccv - Supply the nme cf the origmalentiv mb IS fa Inppenlile meet the LC B. PAPNtrai inoperatle in I or 3 for roaod AND than Condito

1) days from dsccvery of Second failure to Steam Supply meet the LCD lno;erable (continued)

McGuiro Jnits 1 anc 2 3.7.5 I \mondrnent Nzo. 221203

From T.S. 1.3 (Completion Times): Completion Tin-es 1.3

      .0 USE AND APPLICATION
      .3 Completion Times PURPOSE             The purpose of this section is to establish the Completion Time convention and to povide guidance for is use.

BACKORCUND Limiting Conditions for Operaton (LCOsi specify ninimum requirements for ensuring safe opt ration of the unit. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Requ red Actiors) and Completion Time(s). DESCRIPTION The Completion Time is the amount of tVne allowed for completing a Required Action. It is referenced to the lime of discovery of a situation (e.g., iiiupeiable eqiipri rent ui viii iAile i oL wititir i Iii its) Lirat iequiies entering an ACTIONS Condition unless otherwise specified. providing the unit is in a MODE a specified condition stated in the Applicability of the LCO. Required Aclions must e completed prior to the expiration of the speci1iedCoriiAeLiui Time. An ACTIONS CuridiLicit rerirairis iii ellect arid the Requ red Actions apply urtil the Condition no longer exisls or the unit is not within the LCO Applicalility. If situations are discovered that require entry into more than one Condition at a time within a single LCO (niuitiple Conditions), the Required Actions fcc each Condition must be performed within the associated Complelion Time. When in rtultiple Conditions, separate Completion Times are tracked for each Condition starting from the time of discovery of the sitiation that equired entry into the Condition. Once a Condition has been entered. subsequent trains, subsystems, conponents. or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in seporate entry into the Condition, unless specifically stated. The Required Actions cf the Condition continue to apply to each additional failure, with Completion Times based on initial entry into the Condition. However, when a subsequent train, subsystem. coiiponent. or variable expressed in the Condition is discoverec to be inoperable or not within limits, the Completion Time(s) may be extended. To apply this (continued) McGuire Wits 1 and 2 1.3-I Amendment Nos. 1841166

Completion Times 1.3 1,3 ConpIetii Times DESCRIPTION Completion Tire extension, two criteria must first be met. The (eoitinued) S ubsequent inopersbiliti:

a. Must exist concurrent with the first inoperability: and

b. Must remain inoperable or not within limits after the first iiioperability is resolved.

The total Coni pletion Time allowed for completinq a Required Action to address the sLibsequent inoperability shall be limited to the more restrictive of either:

a. The stated Completion Ti me, as nleasLlred from the initial entry into the condition, plus an additional 24 hours; or

b. The stated Completion Ti me as measured from discovers of the subsequent inoperability.

The above Completion Time extensions do not a pply to those Specifications that have exceptions that allo completely separate reentry into the Condition (br each train, subsystem, component, or variable expressed in the Condition) and separate tr:a cking of Completion Times based on this re-entry These exceptions are stated in individual: Specification& The above Completion Time extension does not .apphi to a Completion Time with a modified time zero. This modified time zero may be expressed as a repetitive time (i.e once per 8 hours, where the Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entr) or as a time modified by the phrase from discovers.. Example 1.3-3 illustrates one use of

this type of Completion Time. The 10 day Completion Time specified for Conditions A and B in Example 1.3-3 may not be extended. (continued) McGuilre Units 1 and 2 1.3-2 Amendment Nos. 1B4/166

Completion Times 1.3 1 .3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued) ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A. 1 Restore Function X 7 days Funetion .X train to OPERABLE train status. AND inoperable. 10 days from discovery of failure to meet the LCO B. One 6.1 Restore Function Y 72 hours Function V train to OPERABLE train status. AND inoperable. 10 days from discovery of failure to meet the LCO C. One C. I [Restore Function X 72 hours Function .X train to OPERABLE train status, inoperabl OR AND One C.2 Restore Function V 72 hours Function V train to OPERABLE train status. inoperable. continued) McGuire Units 1 and 2 1.3-6 Amendment Nos. 1841166

Completion Times 13 1.3 Co npletion Times EXAMPLES EXAMPLE 1.3-3 (continued) When one Function Xtrain and one Function Ytrain are inoperable. Condition A and Condition B are concurrently applicable. The Completion Times for Condition A and Condition S are tracked separately for each train starting from the time each train was declared inoperable and the Condition vas entered, A separate Completion Time is established for Condition C and tracked from the time the second train was declared inoperable (i.e., the time the situation described in Condit[on C was discovered). If Required Action Cl is completed. within the specified Completion Time, Conditions S and C are exited. If the Completion Time for Required Action Al has not expired, operation may continue in accordance with Condition A. The remaininq Completion Time in Condition A is measured from the time the affected train was declared inoperable (i.e.. initial entry into Condition A). The Completion Times of Conditions A and S are modified by a ioqical connector with a separate lO day Completion Time measured from the time it as discovered the LCD was not met In this example, without the separate Corn pletion Time, it would be possible to alternate betseen Conditions A, S. and C. in such a manner that operation could continue indefinitely without ever restoring systems to meet the LCD. The separate Com pletion Time modified bc the phrase from discovery of failure to meet the LCD is designed to prevent indefinite continued operation while not meetinq the LCD. This Completion Time allows for an exception to the normal time zero for beqinninq the Completion Time clock In this instance, the Completion Time time zero is specified as commencing at the time the LCD was initially not met, instead of at the time the associated Condition was entered. (continued) McGuire Units 1 and 2 1.3-7 Amendment Nos. 184/166

Parent Question ADMTS02O 1 Pt Given the following conditions:

              -  Unit 1 is operating at 100% power.
              - At 0800 on July 10, during surveillance testing of the Turbine-Driven CA Pump, it is discovered that 1 SA0O49AB, Main Steam Supply from SG 1 B to TD CA Pump, will not open. The steam supply from I B SIG is declared INOPERABLE and Tech Spec 3.7.5 (Auxiliary Feedwater System) is entered at that time.
              -  On July 16, a problem with 1SAOO48ABC, Main Steam Supply from SG 1C to TD CA Pump, is discovered at 0800 hours (valve broken).
              - At 0100 on July 17, the problem with 1SAOO49AB is resolved and the Steam Supply to the Turbine-Driven CA Pump from 1 B S/G is returned to OPERABLE status at that time.

Which ONE (1) of the following states actions required to comply with Technical Specifications 3.7.5 based on the conditions above? A. Restore steam supply to OPERABLE by 0800 on July 19 or place the unit in MODE 3 in 6 hours and MODE 4 in 12 hours. B. Restore steam supply to OPERABLE by 0800 on July 20 or place the unit in MODE 3 in 6 hours and MODE 4 in 12 hours. C. Restore steam supply to OPERABLE by 0800 on July 23 or place the unit in MODE 3 in 6 hours and MODE 4 in 12 hours. D. Restore steam supply to OPERABLE by 0100 on July24 or place the unit in MODE 3 in 6 hours and MODE 4 in 12 hours. Answer 457 B Provide student with copy of T.S. 3.75 and Basis. 0800 7/10 ---- SA49 mop ---- enter 3.7.5 A.1 ---- 7days & 10 day completion

0800 7/16 ---- SA48 mop ---- enter 3.75 B.1 ---- now in 10 day requirement from A.1 ( so due date is 0800 7/20) 0100 7/17 --- SA49 operable ---- still in 10 day requirement from initial entry of 3.7.5 A.1

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 97 GEN2.2 2.2.6 GENERIC Equipment Control

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quipment Control knowledge of the process for making changes to procedures. (CFR: 41.10 / 43.3/45.13) Given the following conditions on Unit 1:

The unit is in MODE 5 preparing for a unit startup after refueling
You are the Unit 1 Control Room Supervisor
A Temporary Test procedure is being run on the 1 B 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 shall be processed as a (1) change.

For any procedure change, a 10CFR5O.59 Evaluation is NOT required (2) Which ONE (1) of the following completes the statements above? A. 1. minor

2. for minor changes ONLY B. 1. major

2. for minor changes ONLY C. 1. minor

2. for minor changes OR if the procedure has been excluded from the 10CFR5O.59 process D. 1. major

2. for minor changes OR if the procedure has been excluded from the 10CFR5O.59 process Tuesday, July 13, 2010 Page 283 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 97 eneraI 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 an evaluation of a procedure may be performed to exclude that procedure from the IOCFR5O.59 process. If the procedure is excluded from the process it is maintained on a list of procedures which are excluded from the IOCFR5O.59 process. Answer A Discussion INCORRECT: See explanation above. PLAUSIBLE: Part 1 is correct. Part 2 is plausible if the applicant does not recall that a procedure may be excluded from the 10CFR5O.59 review process. The answer is partially true in that minor modifications do not require a 50.59 review. Answer B 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 plausible if the applicant does not recall that a procedure may be excluded from the IOCFR5O.59 review process. The answer is partially true in that minor modifications do not require a 50.59 review. Answer C Discussion CORRECT: See explanation above. Answer D Discussion INCORRECT: See explanation above. LAUSIBLE: 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. -- I Basis for meeting the KA The KA 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 10CFR55.43(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 Comprehension NEW Development References Student References Provided Learning Objectives:

1) ADM-OP #2 ferences:

NSD 703 Administrative Instructions for Technical Procedures Tuesday, July 13, 2010 Page 284 of 294

FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 97 ,EN2.2 2.2.6 GENERIC Equipment Control

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uipment Control

 .nowledge of the process for making changes to procedures. (CFR: 41.10 /43.3 /45.13) 401-9 Comments:                                                             RemarkslStatus 401-9 Comments:

Since 50.59 evaluations are NOT ever required for a minor change, I am not convinced that A is plausible. Consider rewriting distractor A or modifying the question stem to validate A. Resolution / Comments: Revised question to ask a separate question about 10CFR5O.59 reviews in general. Also, in the stem of the question, changed should to shall per Lead Examiners General Comments. See attached file for proposed revision to question. Tuesday, July 13, 2010 Page 285 of 294

Question 97

References:

From NSD 703: VERIlY 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 703.7 foe processing a minor procedure change.

3. The list below provides sonic examples of changes that fit the definition of a minor procedure revision or minor procedure change:

Note: This list is not intended to be all inclusive.

a. Incorporating previously approved changes.

b. Correct editonal errors (e.g.. misspe]led words, grammatical errors.

typographical errors).

c. During the certification process, lithe electronic flies have formatting differences from the approved version (e.g.. line endings. page endings. word wraps).

d. Correct, delete, or add information (e.g.. numbering or references to steps.

pages, enclosures, or procedures, work location information, references so other documents. noses, cautions, warnings):

Adding, deletusg. or correcting references to documents that are no longer applicable (e.g., Site Directive deleted in favor of an NSD).
Changing the assigned part sequence numbers but the referenced part does not change (e.g.. Stock Code Numbers).
Changing the part number where an acceptable substitute has been identified under the Acceptable Substitute Program.

C. Correct, delete, or add nontechnical or admiinstrative actions. Add/delete inspection/verification st natures (e.g QC Hold Point Concurrent

                               \eritication).

g. Modify the lormat ot a step or section, but not change the results:

Rephrasing a step. without changing its scope or results, to clarify.
Rephrasing to avoid ambiguous wording (e.g.. nsoving from general to specific).
Changing units in a procedure data sheet (e.g.. Scale bce is marked in inches. but the procedure specified readings in percent. Rather than requiring the technician to perform the conversion each time the procedure is performed. change the procedure to reflect field configuration.)

Reflect changes in administrative work practices that are not commitment iteuss:

Dividing Test Equipment into Test Eqssipnsent and Other Equipment.
Deleting Verify to Control Copy step (now on P PR).
Deleting step for Supenisor to NA steps before beginning procedure (now covered in NSD 704).
Deletuig or adding step to Get key to open cabinet (when cabinet is no longer locked or is being locked).

Document changes initiatedlsubstantiated by other processes evaluated in accordance with NSD 228 (Applicability Detennination):

Equipment nunsber changes evaluated under the engineering change process.

12 REVISION 29 VERifY HARD C:OPY AGAINST WEB SITE IMMEDIATELY PRIOR TO EACH USE

VERilY HARD COPY AGAINST WEB 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 Reviewe( s signature on the Procedure Change Process Record indicates that the procedure has been reviewed in accordance with Appendix M.

5. An Applicability Determination (NSD 228) is NOT required for a minor procedure change.

6. The Reviewer 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 Ensure WARNINGS or CAUTIONS are used appropriately to mininin e risk to personnel or equpment.

c. Ensure the step(s) affected by the change can be accomplished in the sequence written.

d. Ensure the step(s) affected by the change meets the current Technical Specifications, UTSAlt 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 resources.

7. A review of a minor procedure change does not require an extensive review of material not changed. Review the remaining pans of the procedure NOT changed to verily the following:

a. No missing procedure steps or pages.

ii. No obvious formatting problems created by the revision:

lnappropnate page breaks
Cautions. Wamings or Notes not on same page with applicable step
Incorrect step numbering

c. Change was appropriately incorporated into ALL affected parts of the procedure.

S. The Reviewer shail determine the need for cross-disciplinary or additional reviews based on the following:

a. Response of a system -under direct conirol of another group will be altered-.

Steps in a procedure may affect the use or operation of equipment under another s control. 1 group

c. Another group will be required to provide personnel to assist in the performance of a procedure.

ci. 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 which involve Environmental Emergency Response plans (part of the Site Emergency Plan) shall also be reviewed by the Site Environmental Management Section.

26 REVISION 29 VERIFY HARD COPY AGAINST WEB SITE LSLMEDL4JELY PRIOR TO EACH USE

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 98 2598 GEN2.3 2.3.12 GENERIC Radiation Control

                      -             -

adiation Control

 .nowledge 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: 41.12/45.9 / 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. Tuesday, July 13, 2010 Page 286 of 294

______ _____________________ _____ ________ ______ FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 98 2598 General Discussion his 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, 10CFR 100 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 e 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 (10CFRIOO) and the limiting condition is a condition of a SRO license required to operate the on._ -__ -__ 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 Questionsi dated 03/I 1/2010) under the Screening Criteria for question linked to 10CFR55.43(b)(2) (Tech Specs):

1) It can NOT be answered solely by knowing < 1 hour 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 SLCI6.l1.20 GEN2.3 2.3.12 GENERIC Radiation Control

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liation Control owledge 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 / 45.9 / 45.10) Tuesday, July 13, 2010 Page 287 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE

2010 MNS SRO NRC Examination QUESTION 98 O1-9 Comment Remarks!Status 401-9 Comments: No comment. Resolution I Comments: Tuesday, July 13, 2010 Page 288 of 294

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 WEWGOO7 x

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-i 33).

APPLICABILITY At all times. REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Quantity of radioactive A.i Suspend all additions of Immediately material in tank not radioactive material to the within limit, tank. AND A.2 Reduce the tank contents 48 hours to within limit. TESTING_REQUIREMENTS TEST FREQUENCY TR 16.1 1.20.1 Verify the quantity of radioactive material contained in 24 hours 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 WEWGN04 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

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 99 259 GEN2.3 2.3.14 GENERIC Radiation Control

                  -             -

adiation Control

 .nowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. (CFR: 41.12 I 43.4 / 45.10)

Given the following plant conditions:

You are tasked to evaluate four available work teams to perform repairs in a 1500 mREM/hr radiation field Which ONE (1) of the following work teams would maintain station radiation dose ALARA?

A. Two qualified male workers can complete the task working together in 15 minutes. Each worker has accumulated 325 mREM for the year. B. A team consisting of a qualified male and qualified female worker can complete the task working together in 20 minutes. Each worker has accumulated 100 mREM this year. C. A qualified male worker who has previously performed this task can complete the task in 20 minutes. However, he has exceeded his Alert level for exposure and will require a dose extension. D. A team consisting of a qualified declared pregnant female worker and a non-qualified male worker who needs to qualify on this task can complete the task working together in 1 5 minutes. The female has no dose and the male worker has accumulated 200 mREM for the year. Tuesday, July 13, 2010 Page 289 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 99 General Discussion To maintain station dose ALARA. the worker/team with the lowest dose for the job Consistent with meeting all other exposure limits should be selected. The single qualified male worker would receive 500 mREM to perform the work. Although he would exceed his annual ADMINISTRATIVE limit, his total exposure would be less than the other teams selected for the work. - Answer A Discussion See explanation above. rRRECT PLAUSIBLE: It would be reasonable for the applicant to eliminate the single male worker since he would exceed an ADMINISTRATIVE dose limit, even though the total dose would be less (500 mREM). However, this is the Correct choice. The worker would not exceed a I OCFR2O dose limit. The team consisting of the delcared pregnant female and male worker would have a dose of 750 mREM total and their total dose for the year would be less than the two male workers. The female worker would not exceed the total dose limit for a declared pregnant female worker (500 mREM). However, the female worker would exceed the monthly dose limit for a declared pregnant female (50 mREM). Therefore, this team could not be used. The team consisting of the male worker and female worker would receive a total dose of 1000 mREM which would be higher than the dose for the two male workers working together (750 mREM). It is therefore plausible for the applicant to conclude that the team consisting of the two male workers would be the correct choice. Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: It would be reasonable for the applicant to eliminate the single male worker since he would exceed an ADMINISTRATIVE dose mit. even though the total dose would be less (500 mREM). However, this is the correct choice. The worker would not exceed a I OCFR2O dose nit. The team consisting of two male workers would receive less total dose (750 mREM). However, it is plausible for the applicant to conclude that the team consisting of the male and female worker would be a better choice since their total dose for the year would be less. The team consisting of the declared pregnant worker and the male worker would have less total dose (750 mREM). However, the female worker would exceed the monthly dose limit for a declared pregnant female worker (50 mREM). Answer C Discussion CORRECT: See explanation above. Answer D Discussion nORRECT: See explanation above. PLAUSIBLE: It would be reasonable for the applicant to eliminate the single male worker since he would exceed an ADMINISTRATIVE dose limit, even though the total dose would be less (500 mREM). However, this is the correct choice. The worker would not exceed a I OCFR2O dose limit. It would be reasonable for the applicant to eliminate the team consisting of the male and female worker since the total dose for the job would be higher for this team 1000 mREM than is would for the team consisting of the declared pregnant female and male worker (750 mREM). The total dose for the job would be the same for the team consisting of the two male workers and the team consisting of the declared pregnant female worker and the male worker (750 mREM). However, the annual dose for the two male workers would be higher and it would therefore be plausible for the applicant to chose the team conisting of the declared pregnant female and male workers. Basis for meeting the KA

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The KA is matched because the applicant must evaluate the radiation hazard to a team of workers performing repair activities. 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 dose limits from mory and it requires the applicant to calculate the total dose for each team and compare them to each other. -

 ,asis for SRO only This question is SRO level knowledge because it can not be answered solely by RO knowledge of radiological safety principles (e.g., RWP requirements, stay-time, DAC-hours, etc.).

Tuesday, July 13, 2010 Page 290 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 99 2599 jquires the applicant to analyze the makeup of the available repair teams with regards to the total exposure for the job and the accumulated (4 iual exposure for the teams and make a determination as to which team would be the correct choice to maintain station dose ALARA. Job Level Cognitive Level QuestionType Question Source SRO Comprehension BANK MNS Exam Bank Q# RADRPNO18 Development References Student References Provided Lesson Plan OP-MC-RAD-RP Lesson Plan Objective RAD-RP #135 GEN2.3 2.3.14 GENERIC Radiation Control

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Radiation Control Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities. (CFR: 41.12/ 43.4 / 45.10) 4O1-9 Comments: RemarkslStatus - Proposed replacement for 20l0 NRC Q99. Replacement question approved RFA 07/08/10. Wednesday, July 14, 2010 Page 291 of 294

Overlap Replacement Question 99: Objective #132 Prior to the use of the Emergency Exposure dose limits, the following approvals ( written or verbal are

required:

Radiation Protection Manager or designee
Emergency Coordinator or EOF Director The person(s) who is/are to receive the dose must sign that they have been informed of the potential dose they will receive, have been fully briefed on the task to be accomplished and the risks of this exposure.

Objective #134 Regulatory Guide 8.14 requires a personnel neutron dosimeter if the neutron dose equivalent is likely to exceed 100 mrem in a quarter. Duke Power has an administrative requirement which requires all personnel entering the RCA to wear a TLD (which measures neutron dose equivalent). Estimation of neutron exposure is a method used to temporarily track exposure until the TLD is processed. Estimated neutron exposure tracking for personnel is required if the neutron dose equivalent is likely to exceed 10 mrem per entry or per job if consecutive multiple entries are required. There are two methods used to estimate neutron exposure:

One method is to measure the neutron dose rate and then calculate the exposure based on stay time.
The second method is to determine the gamma exposure dose and neutron exposure dose for the given area. If it is determined that the neutron to gamma ratio is essentially constant during the period of personnel exposure, then a gamma/neutron ratio can be utilized. The gamma dose received can be ratioed to find the neutron dose received.

Objective #135 ALARA is a philosophy aimed at the minimizing exposure thru a management commitment. The goals and efforts of the McGuire Nuclear station Program are simple:

To maintain the annual dose to each individual ALARA
To maintain the collective dose (total person-Rem ) ALARA
Both points have to be considered simultaneously, as one without the other is not ALARA.

Question 99 Parent Question: RADRPNO1 8 1 Pt(s)As an SRO working on a Complex Maintenance Plan you are asked to evaluate four possible work teams who must repair filter housing in a 1500 mRem/hr radiation field. Which one of the following work teams would maintain station ALARA? A. A qualified male worker who has previously performed this task. He can complete this job in 20 minutes. This worker has exceeded his Alert level for exposure and will require a dose extension. B. Two male workers who are qualified to perform the task. Together they can perform the task in 15 minutes. Both workers have already accumulated 325 mRem this year. C. A team of a female worker who is qualified to perform the task and a male worker who needs to qualify to this task. The female is a declared pregnant worker. The team will need 15 minutes to complete the task. The female has no dose and the male worker has 200 mRem for the year. D. A team of a male and female both are qualified to the task but will take 20 minutes to complete the task. Each has less than 100 mRem this year. Answer 317 A

Distracter Analysis: A. Correct: 500 mR total B. Incorrect: 750 per mrem total Plausible C. Incorrect: Declared pregnant worker. Plausible: D. Incorrect: 1000 mrem total Level: SRO KA: G2.3.2 (2.5/2.9) Lesson Plan Objective: RAD RP Ojb. 135 Source: New Level of knowledge: comprehension

References:

1. OP-.MC-RAD-RP page 73

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 100 GEN2.4 2.4.40 GENERIC Emergency Procedures / Plan

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ergency Procedures I Plan

 ..nowledge of SRO responsibilities in emergency plan implementation. (CFR: 41 10 / 43.5 / 45.11)

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 hour

2. hour B. 1. 1 hour

2. 4 hours C. 1. 15 minutes

2. hour D. 1. 15 minutes

2. 4 hours Tuesday, July 13, 2010 Page 292 of 294

_______ __________ __________ FOR REVIEW ONLY - DO NOT DISTRIBUTE 2010 MNS SRO NRC Examination QUESTION 100 2600

    *eneral Discussion In 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, RP/29 the follow-up notification requirement will change from 4 hours (NOUE) to a new requirement of 1 hour for the new Alert classification. The NRC fication procedure, RP/1 0 requires that the NRC be notified immediately but not more than 1 hour after a change in classification.

Answer A Discussion CORRECT: See explanation above Answer B Discussion INCORRECT: See explanation above. PLAUSIBLE: Part I is correct and therefore plausible. Part 2 is plausible because the follow notification requirement for a NOUE is 4 hours 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 rfiCORRECT: See explanation above. PLAUSIBLE: Part I 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. irt 2 is plausible because the follow notification requirement for a NOUE is 4 hours 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 implementin g the Emergency Plan (i.e. ptification 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 understandin g the rules pertaining to offsite notification and apply them to a dynamic situation. Basis for SRO only This question is not tied to 10CFRSO.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 M14S it is the responsabilit y 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 References Provided Learning Objective: l)EP-EMP#12 and #14 (Notifications to Offsite Agencies From The C/R)

c. 4.2 Pg 1 of 8 RP/10 (NRC Immediate Notification Requirements)

Enc4.lPglofl4 Tuesday, July 13, 2010 Page 293 of 294

FOR REVIEW ONLY DO NOT DISTRIBUTE - 2010 MNS SRO NRC Examination QUESTION 100 2600 GEN2.4 2.4.40 - GENERIC Emergency Procedures / Plan

ergency Procedures / Plan nowledge of SRO responsibilities in emergency plan implementation. (CFR: 41.10 / 43.5 / 45.11) 401 -9 Comments: RemarkslStatus L 40 1-9 Comments: No comment. Resolution / Comments: N/A Tuesday, July 13, 2010 Page 294 of 294

Question 100

References:

OP-MC-EP-EMP Obj: 12 & 14 OBJECTIVES NNLLL OBJECTIVE L L P P 0 OORS R ROO 10 Given the EPIPs and the emergency situation, classify the event.

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Given the EPIPs and the emergency situation, provide the ii appropriate Protective Action Recommendations (PARs). x x 12 Given the EPIPs and the emergency situation, complete the Emergency Notification Form. X X Describe the use of the Selective Signaling Phone System to 13 notify the State and County. x x x x x 14 Given the EPIPs and the emergency situation, complete the appropriate portions of the procedure for an NRC Event X X Notification Worksheet. 15 State the requirements for Initial and Follow-up Notifications including: X X X

      . Time Requirements
      . Agencies to be contacted

From 0P101B157001029 Enc. 4.2 Pg lof 8 Enclosure 4.2 RP/O/B15700/029 Completion and Transmission of a Page 1 of S Follow-up Message NOTE: New initial messages for higher classification upgrades are addressed in Enclosure U (PIP M-O1-37l 1} L Make follow-up notifications according to the following table: Follow-up Notifications L Follow-up notifications to the State(s) and Counties must be nude accord ing to the following schedule:

      -For a NOUE. every 4 hours until the emergency is terminate& For ALER T. SAE. or GE even 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 individ ual agencv Documentation shall be maintained for any axeed upon schedule changa The interval for ALERT, SAE. and GE shall not be greater than 2 hours. 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 ENE In this case, the offsite agencies must be notifie d that the pending follow-up is being superseded by an upgrade to a higher classification and inform ation will be provided. 3 Follow-up messages in the General Emergency classification that involv e an upgrade in PARs must be communicated to the offsite agencies as soon as possible and within 15 minutes 1 Complete an Emergency Notification Form by one of the following: C Obtain a preprinted ENE OR Q Obtain a blank ENF.

From 0P101B157001010 Enc. 4.1 Pg lof 14 Enclosure 4.1 RP!OJA5700/O1O Events Requiring NRC Notification Page 1 of 14 41.1 Events RequirinafldidEDLUE NOTIFICATIONS: REPORTING TIME REQUIRSIENFS REPORTABlE EVENIS Correspondino 10CFR Sectionin Brackets 1] 4.1.1.1 The declaration of any of the Emergency Classes 4.1.1.1 Iinmediatelvaflernotiflcationto state(s) andlocal roverument(counties) specified m the McOuae Emergency Plan and not later than one hour after the time the Emergency CIa sins [50.72a(l)(i)) declared. Immediately report any change from one Emergency Class to another or a termination of the Emergency Class. (Use Enclosure 4.2,) or See thilowup requirements in section 4.1.6 of this procedure. 150.72c(I Xii)) any chanze from one Emernencv Class to another or f50.72c(,lXiii)l a termination ofthe Emergency Class 4.1.1.2 Events mvolvinrecemng amlopeningpackages 4.1.1.2 NOTE Reportingsimer 10CFR2O.1906 shauldbemade asfollaws: the containing quantities of radioactive material in excess of licensee shall immediately not4 the final delivery carrier and by 120.1906) a Type A quantity as defined in section 71.4 and telephone and teleaan mailgrani. or thcsimile and the NRC Operations appendix A to part 71 of this chapter when; Center at 9-301-816-5100. 120.1906d0) 1) Removable oadaoactave suffice contamination exceeds the limits of section 71.87(1) of 10CFR2O; 01 [20.l906d(2)) 2) External radiation levels exceed the limits of section 71.47 of this chapter. 1.1.1.3 Any lost, stolen. ormissing licensedmatenaim an 4.1.1.3 I&ate1yaflerits occurrencebecomeshownto thelicensee. agegate quantity equal to or eater than 1,000 times [10.2201a(i)) the quantity specified in appendix C to part 20 under such circumstance that it appears to the licensee that an eaposure could result to persons m unresticted areas. 01 [20.2201a(ia)] 3) \Vithin3O days after the occusrence of any lost stolea or mission licensed matenal becomes known to the licensee, all licensed material in a quantity fleater than 10 times the quantity specified in appendix C to part 20 that is still missiur at this time.}}

ML102640737

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1.0 INTRODUCTION

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