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.JS Sim Phase USE THIS > Safety Analysis (Drey) > Flashcards

Safety Analysis (Drey) Flashcards

1
Q

The PVGS safety limit for DNBR is ___?

A. < 1.24

B. > 1.24

C. < 1.34

D. > 1.34

A

D. > 1.34

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2
Q

Which one of the following would be considered a Single Failure in our Accident Analysis?

A. A RCS Pipe Rupture

B. A Failure of Two Diesels to Start

C. A dropped or slipped single CEA

D. A Turbine Trip with a Loss Of Offsite Power

A

D. A Turbine Trip with a Loss Of Offsite Power

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3
Q

A ______________ is the event or events in a category of events that would require LCOs, procedures, plant design, etc. to minimize the consequences to the health and safety of nuclear workers and the public.

A. design fault

B. limiting fault

C. limiting event

D. design accident

A

C. limiting event

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4
Q

When describing an ATWS event, which of the following is a ‘transient of greatest concern’ ?

A. Small break LOCA

B. ESD inside containment

C. Dropped or Slipped CEA

D. Loss of Normal Electrical Power

A

D. Loss of Normal Electrical Power

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5
Q

Which of the following systems is designed to protect the plant in the event of an ATWS?

A. RPS

B. AFAS

C. SBCS

D. DAFAS

A

D. DAFAS

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6
Q

Given the following conditions:
• Unit 1 is at 100% power when a loss of offsite power occurs (LOOP).
• Both DGs fail to start.

Which of the following describes how the safety analysis assumes the reactivity control safety function is maintained?

A. Plant cooldown is limited to 75°/hr.

B. The plant is maintained in hot standby.

C. Within 1 hour RCS pressure is reduced below SIT pressure.

D. After verification of all CEAs inserted, the plant will be rapidly cooled to SDC entry conditions.

A

B. The plant is maintained in hot standby.

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7
Q

What is the purpose of RCP coastdown following a turbine trip with a LOOP?

A. Prevents RCP damage.

B. Helps establish natural circulation.

C. Prevents potential voltage spike from disabling vital buses.

D. Provides for maximum decay heat removal immediately following the trip.

A

D. Provides for maximum decay heat removal immediately following the trip.

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8
Q

For which of the following accidents does our Chapter 15 analysis assume a concurrent loss of offsite power (LOOP)?

A. RCS dilution.

B. Seized RCP Rotor.

C. Uncontrolled CEA withdrawal at power.

D. Steam Generator Tube Rupture (SGTR).

A

D. Steam Generator Tube Rupture (SGTR).

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9
Q

The UFSAR assumes that the release to the atmosphere following the limiting SGTR is terminated _______ hours after the event.

A. 1

B. 4

C. 8

D. 24

A

C. 8

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10
Q

Given the following conditions:

  • Unit 1 was operating at 100% power when the Reactor trips on low Pressurizer pressure.
  • A SIAS and CIAS automatically initiated.
  • A Loss of Offsite Power occurred when the Main Turbine tripped.

After completion of SPTA the following conditions exist:

  • Reactor Power is lowering.
  • Pressurizer level 13% and slowly lowering.
  • Pressurizer Pressure is 1810 psia and stable.
  • Subcooling is 40ºF and slowly degrading.
  • S/G pressures are 1180 psia and slowly rising.
  • TC is 565ºF and slowly rising.
  • Feed flow to both Steam Generators are equal at 500 gpm.
  • Steam Generator ‘1’ level is 10% NR and slowly rising.
  • Steam Generator ‘2’ level is 30% NR and rising rapidly.

Based on the given indications what Time Critical Operator Action is credited in 40DP-9ZZ04, Time Critical Action Program?

A. Both Steam Generator tubes are covered within 46 minutes.

B. Any release is terminated after 16 hours when MODE 5 is achieved.

C. Within 70 minutes affected Steam Generator isolated (Hot leg target temperature reached).

D. After 21.5 minutes a RCS cooldown is performed with only Steam Generator 1 ADVs at approximately 100ºF/hr to TH = 540ºF.

A

C. Within 70 minutes affected Steam Generator isolated (Hot leg target temperature reached).

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11
Q

The UFSAR analysis for a SGTR at Palo Verde assumes that the failures for the SGTR includes a loss of offsite power and the added single failure of . . .

A. RCS iodine spike.

B. of a Main Steam Safety Valve open.

C. a stuck open ADV on the affected SG.

D. an Auxiliary Spray Valve to operate with the other out of service for maintenance.

A

C. a stuck open ADV on the affected SG.

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12
Q

During a Loss of Main Condenser Vacuum event, the safety analysis assumes the plant is shutdown by the…

A. VOPT trip.

B. low SG level trip.

C. high SG pressure trip.

D. high PZR pressure trip.

A

D. high PZR pressure trip.

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13
Q

All assumptions in the UFSAR analysis for Decrease in Heat Removal are aimed toward…

A. minimizing DNBR.

B. maximizing RCS pressure.

C. maximizing linear heat rate.

D. maximizing reactor power peaking.

A

B. maximizing RCS pressure.

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14
Q

What is the most limiting event with regards to Decrease in Heat Removal TAA.

A. Loss of all Feed

B. Main Turbine Trip

C. RCS and SG heat up

D. Loss of Condenser Vacuum

A

D. Loss of Condenser Vacuum

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15
Q

Per the Technical Specifications bases for LCO 3.7.5, Auxiliary Feedwater (AFW) System, what are the limiting Design Basis Accidents (DBAs) and transients for the AFW System?

  1. Large Break LOCA
  2. Feedwater Line Break
  3. Main Steam Line Break

A. 1 ONLY

B. 2 ONLY

C. 1 and 3 ONLY

D. 2 and 3 ONLY

A

D. 2 and 3 ONLY

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16
Q

Which of the following is identified as the parameter of concern for a Decrease in RCS Flow fault?

A. ASI

B. LHR

C. DNBR

D. Peak centerline temperature

A

C. DNBR

17
Q

Which of the following describes the analysis assumed trip mechanism for a Reactor Coolant Pump shaft break?

A. High LPD

B. Low DNBR

C. Low S/G delta P

D. High pressurizer pressure

A

C. Low S/G delta P

18
Q

Which of the following initial conditions has the most impact on the LOCA safety analysis?

A. ASI

B. RCS pressure

C. RCS temperature

D. Reactor power history

A

D. Reactor power history

19
Q

Which of the following DOES NOT describe one of the 5 phases of a Large Break LOCA for the FSAR analysis.

A. Refill

B. Blowdown

C. Post Reflood

D. Pre Reflood

A

D. Pre Reflood

20
Q

According to the acceptance criteria for the ECCS, established by 10CFR50.46, “the calculated total oxidation of the cladding shall nowhere exceed % of the total cladding thickness before oxidation.”

A. 1

B. 17

C. 27

D. 35

A

B. 17

21
Q

ECCS design criteria ensures that nowhere shall the clad oxidation exceed 17%. The reason behind this is:

A. The zirc water reaction will be stopped by the oxide layer.

B. Clad failure from increased gas gap pressure will be minimized.

C. Clad damage or failure from increased peak centerline temperature will be minimized.

D. Sufficient clad material strength will be maintained to prevent brittle fracture if cold ECCS hits the clad.

A

D. Sufficient clad material strength will be maintained to prevent brittle fracture if cold ECCS hits the clad.

22
Q

In respect to the ECCS acceptance criteria, fuel element cladding temperature should be maintained below 2200 degrees in order to…

A. prevent clad ballooning that could lead to a possible failure.

B. prevent the zirc-water reaction from becoming self-sustaining.

C. prevent the onset of zirc-water reaction within the core.

D. limit the clad failure when cold water is introduced during a LOCA.

A

B. prevent the zirc-water reaction from becoming self-sustaining.

23
Q

When describing the Emergency Core Cooling System (ECCS) acceptance criteria, fuel element cladding temperature should be maintained below 2200 0F in order to…

A. prevent the onset of zirc-water reaction within the core.

B. limit the clad failure when cold water is introduced during a LOCA.

C. prevent the exothermic zirc-water reaction from becoming self-sustaining.

D. prevent H2 from being produced which could lead to an explosive containment.

A

C. prevent the exothermic zirc-water reaction from becoming self-sustaining.

24
Q

Which one of the following is the worst case following a large break LOCA?

A. No ECCS failures occurring.

B. A single Emergency diesel generator failing.

C. A single LPSI pump failing.

D. A single HPSI pump failing.

A

A. No ECCS failures occurring.

25
Q

Which of the following is the number of Safety Injection Tanks that are ASSUMED to inject their water into the core on the design basis large break LOCA in the Cold leg?

A. 1

B. 2

C. 3

D. 4

A

C. 3

26
Q

Assuming that the design large break LOCA is occurring, how much HPSI injection is assumed to reach the core in the Safety Analysis?

A. 50% of one HPSI pump

B. 50% of two HPSI pumps

C. 75% of one HPSI pump

D. 75% of two HPSI pumps

A

D. 75% of two HPSI pumps

27
Q

All of the following are required manual operator actions following a LOCA except for:

A. Providing long-term cooling via the HPSI hot leg injection valves if required.

B. Closing the LPSI and HPSI recirc valves after a RAS.

C. Closing the RWT discharge valves after a RAS.

D. Cooling down the SGs using auxiliary feed and ADVs as required.

A

B. Closing the LPSI and HPSI recirc valves after a RAS.

28
Q

If a letdown line break were to occur outside CTMT, the results of the safety analysis indicate that…

A. DNBR drops below the value at which the fuel pins would be calculated to experience DNB.

B. DNBR stays above the value at which the fuel pins would be calculated to experience DNB.

C. Operator action is NOT required to isolate letdown during this event.

D. Automatic isolation of letdown prevents primary system coolant from being released to the auxiliary bldg.

A

B. DNBR stays above the value at which the fuel pins would be calculated to experience DNB.

29
Q

What is the single failure assumption of the safety analysis associated with the most limiting reactivity anomaly event?

A. LOP

B. Stuck ADV

C. Stuck CEA

D. Failure of a HPSI to start

A

A. LOP

30
Q

Which one of the following is the operator action assumed in the Transient Accident Analysis (if any) that terminates the approach to DNB Ratio on a Single Full Strength CEA Drop? Assume the rod is inserted to the bottom.

A. Take action to reduce power.

B. No operator action is analyzed.

C. Trip the plant and enter SPTAs.

D. Declare COLSS inoperable and enter applicable LCO condition.

A

A. Take action to reduce power.

31
Q

Which of the following describes the mitigating action assumed in the safety analysis for a Single Full Strength CEA Drop?

A. Restore the dropped CEA to group height.

B. Reduce Turbine load to restore Tcold to program.

C. Commence a power reduction within 10 minutes.

D. Commence a power reduction within 15 minutes.

A

D. Commence a power reduction within 15 minutes.

32
Q

Which of the following events, that are considered reactivity anomalies in the UFSAR, is considered the most limiting fault?

A. CEA drop.

B. CEA ejection.

C. An idle RCP start.

D. Inadvertent dilution.

A

B. CEA ejection.

33
Q

Which of the following describes how CETs can be used to assess core damage?

Can be used to determine . . .

A. the percentage of core melt during a cladding failure.

B. the I-131 release rate to the RCS during the a clad rupture.

C. the amount of core damage during a core uncovery event.

D. subcooled conditions when vessel temperatures exceed 2300ºF.

A

C. the amount of core damage during a core uncovery event.

34
Q

During Accident Conditions resulting from a Design Basis Earthquake, which system provides more reliable indications and why?

A. QSPDS, it is seismically qualified.

B. ERFDADS, its inputs are electrically isolated from non safety related inputs.

C. Plant Computer, it has redundant inputs powered from multiple class power supplies.

D. SPDS, it has required STs to ensure its OPERABILITY during all post accident conditions.

A

A. QSPDS, it is seismically qualified.

35
Q

During the blowdown stage of a LOCA event, which of the following is the most significant source of hydrogen generation?

A. Radiolysis

B. Steel - steam reaction

C. Zirconium - steam reaction

D. Corium - concrete interaction

A

C. Zirconium - steam reaction

36
Q

If the hydrogen concentration inside the containment reached 22%, which of the following statements would apply?

A. The mixture is too rich to burn.

B. The mixture is too lean to burn.

C. The mixture is in the detonable region.

D. The mixture could burn without producing a pressure spike.

A

C. The mixture is in the detonable region.

37
Q

Which of the following initial conditions maximizes the potential for fuel failure in the event of a LOCA?

A. Time in core life.

B. Initial RCS pressure.

C. Power history of core.

D. Initial RCS temperature.

A

C. Power history of core.

38
Q

Given the following plant conditions:

  • Unit 1 is at 100% power when an RCS leak is in progress
  • 40AO-9ZZ02, RCS Leakage, has been implemented and conditions exist requiring a reactor trip
  • No PPS parameters have been exceeded when the UNIT CRS directs a manual reactor trip
  • The operator depresses all four reactor trip pushbuttons and the reactor does not trip

Based on these actions an ATWS has…

A. occurred and emergency boration must be performed immediately.

B. NOT occurred and SPTAs should be entered immediately.

C. occurred and SPTAs must be entered immediately.

D. NOT occurred and emergency boration must be performed immediately.

A

B. NOT occurred and SPTAs should be entered immediately.

39
Q

What is the most limiting event with regards to Decrease in Heat Removal TAA?

A. Loss of all Feed

B. Main Turbine Trip

C. RCS and SG heat up

D. Loss of Condenser Vacuum

A

D. Loss of Condenser Vacuum

.JS Sim Phase USE THIS (19 decks)

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