ML072970101

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July-August Exam 50-325, 324/2007301 Final Simulator Scenarios (Scenario 3 of 4) (Section 5 of 5)
ML072970101
Person / Time
Site: Brunswick  Duke Energy icon.png
Issue date: 01/31/2007
From:
- No Known Affiliation
To:
Office of Nuclear Reactor Regulation
References
2APP-A-03, 50-324/07-301, 50-325/07-301 50-324/07-301, 50-325/07-301
Download: ML072970101 (9)
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Text

Unit 2 APP A-03 2-10 Page 1 of 1 RHR HX B OUTLET HI CONDUCTIVITY AUTO ACTIONS NONE CAUSE

1. High conductivity water at the outlet of RHR Heat Exchanger 2B due to crud buildup or a service water leak (10 ~o/cm).
2. Defective Conductivity Cell, E11-CE-N001B.
3. Defective Conductivity Indicating Switch, E11-CIS-R001B.
4. Circuit malfunction.

OBSERVATIONS

1. Conductivity indication greater than 10 ~o/cm (E11-CIS-R001B on Instrument Rack H21-P021 in the south RHR pump room) .

ACTIONS

1. Direct E&RC to sample the outlet of RHR HX 2B.
2. If the sample of RHR HX 2B outlet conductivity is less than 10 ~o/cm, omit the rest of these steps.
3. If the-RHR System is operating in the shutdown cooling mode and the sample of the RHR HX 2B outlet conductivity is greater than 10 ~o/cm, shift shutdown cooling to Loop A or an alternate mode per OP-I7, Residual Heat Removal System or OP-32, Condensate and Feedwater System.
4. If the RHR System is operating in the supplemental fuel pool cooling mode, shut down Loop B of the RHR System per OP-17 and operate the Fuel Pool Cooling System per OP-I3, Fuel Pool Cooling System, to maintain fuel pool temperature.
5. If a tube leak or circuit malfunction is suspected, ensure that a WR/JO is prepared.

DEVICE/SETPOINTS Conductivity Indicating Switch E11-CIS-R001B 10 ~o/cm POSSIBLE PLANT EFFECTS

1. Possible fouling of fuel cladding surfaces and stress corrosion of components due to the introduction of service water into the vessel or suppression pool.
2. High conductivity may result in a Technical Requirements Manual compensatory measure.

REFERENCES

1. LL-9364 - 57
2. Technical Requirements Manual TRMS 3.13
3. OP-17, Residual Heat Removal System
4. OP-13, Fuel Pool Cooling System
5. OP-32, Condensate and Feedwater System 12APP-A-03 Rev. 44 Page 37 of 100 I

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  • SERVICE WTR EFFLUENT RAD HIGH Unit 2 APP UA-03 5-5 Page 1 of 1 AUTO ACTIONS NONE CAUSE
1. High radiation level in the service water effluent caused by inleakage from one or more of the following:
a. RHR Heat Exchanger 2A.
b. RHR Heat Exchanger 2B.
c. RBCCW Heat exchangers.
d. Division I RHR pump seal coolers.
e. Division II RHR pump seal coolers.
2. Circuit malfunction.

OBSERVATIONS

1. Service water effluent radiation monitor indicates high on Panel H12-P604.
2. Increasing radiation level indicated on Recorder D12-R604.

ACTIONS

1. Enter EOP-04-RRCP, Radiological Release Control, and execute concurrently with this procedure.
2. Identify and isolate the source of inleakage.
3. If a circuit malfunction is suspected, ensure that a Trouble Tag is prepared.

DEVICE/SETPOINTS D12-RM-K607C K6 Relay Variable POSSIBLE PLANT EFFECTS

1. Possible release to the environs.
2. An aDCM Required Compensatory Measure may exist.
3. This annunciator is required to be operable to support Service Water Effluent Rad Monitor operability; annunciator inoperability will result in a Required Compensatory Measure.

REFERENCES

1. LL-9353 - 36
2. EOP-04-RRCP
3. aDeM 7.3.1, Table 7.3.1-1 Function 3 and 7.3.3

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  • 1.0 1.1 SYMPTOMS AREA RAD REFUEL FLOOR HIGH (UA-03 3-7) is in alarm.

1.2 AREA RAD NEW FUEL STORAGE HIGH (UA-03 4-7) is in alarm.

1.3 PROCESS RX BLDG VENT RAD HI (UA-03 4-5) is in alarm.

1.4 TURB BLDG VENT RAD HIGH (UA-03 3-3) is in alarm.

1.5 Area Radiation Monitor (ARM) is in alarm.

1.6 Continuous Air Monitor (CAM) is in alarm.

1.7 Turbine Building once-through effluent monitor indicates elevated activity.

1.8 Routine sUNeys indicate high radiation, contamination and/or airborne activity.

1.9 Report of spill, leak, or potential damage to new or spent fuel.

2.0 AUTOMATIC ACTIONS 2.1 IF PROCESS RX BLDG VENT RAD HI-HI (UA-03 3-5) is in alarm, THEN the following actions occur:

Reactor Building Ventilation isolation 0 SBGTS auto start 0 Group 6 Isolation. 0 3.0 OPERATOR ACTIONS 3.1 Immediate Actions

[ill 3.1.1 IF a fuel assembly was dropped or damaged, THEN 0 ENSURE the Control Room Emergency Ventilation System (CREVS) is in operation .

  • IOAOP-05.0 Rev. 20 Page 2 of 10 I
  • 3.0 OPERATOR ACTIONS 3.2 Supplementary Actions 3.2.1 EVACUATE unnecessary personnel from the affected area.

o NOTE: Simultaneous or multiple actuations of fire alarms within the Reactor Building may provide additional indication of a High Energy Line Break (HELB).

NOTE: The Reactor Building Sprinkler System is required to be isolated within 15 minutes of indication of a HELB. Location of the system isolation valves has been provided below to expedite isolation:

FP-PIV45, Near Radwaste Building - Northeast Corner FP-PIV33, East end of the Unit 2 Reactor Building FP-V214, Deluge Valve Pit NO.1 next to Unit 1 Reactor Building FP-V214, Deluge Valve Pit NO.2 next to Unit 2 Reactor Building

  • 3.2.2 1.

IF a HELB is indicated in the Reactor Building, THEN PERFORM the following:

UNLOCK AND CLOSE UNIT 1(2) REACTOR BUILDING SPRINKLER SHUTOFF VALVE, 2-FP-PIV45(33).

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2. IF additional sprinkler system isolation is required, THEN o UNLOCK AND CLOSE DELUGE VAL VE 1(2)FP-DV20 SHUTOFF VALVE, 1(2)FP-V214.
  • IOAOP-05.0 Rev. 20 Page 3 of 10 I

3.0 OPERATOR ACTIONS

4. IF fuel failure has occurred, THEN ISOLATE the following systems as necessary:

Reactor Water Cleanup D

- Main Steam Lines (MSIVs) D RHR Shutdown Cooling Mode. D 3.2.12 WHEN directed by E&RC, THEN normal access may be D granted to areas evacuated.

10AOP-OS.0 Rev. 20 Page 7 of 10 I

STEPS PC/P-14 through PC/P-16 CONSIDER ANTICIPATION OF EMERGENCY DEPRESSURIZATION PER RC/P SECTION OF "REACTOR VESSEL CONTROL PROCEOURE~ (EOP RVCP)

PSP PCIP-15 PCIP-16 STEP BASES:

If suppression pool and/or drywell sprays could not be initiated or if operation was not effective in reversing the rising trend of primary containment pressure, as evidenced by not being able to maintain suppression chamber pressure below the Pressure Suppression Pressure, the reactor is depressurized to minimize further release of energy from the reactor vessel to the primary containment. This action serves to terminate, or reduce as much as possible, any continued primary containment pressure rise.

The Pressure Suppression Pressure is defined to be the lesser of either (1) the highest suppression chamber pressure which can occur without steam in the suppression chamber airspace or (2) the highest suppression chamber pressure at which initiation of reactor depressurization will not result in exceeding Primary Containment Pressure Limit A before reactor pressure drops to the Minimum Reactor Flooding Pressure, or (3) the highest suppression chamber pressure which can be maintained without exceeding the suppression pool boundary design load if SRVs are opened. This pressure is a function of primary containment water level, and it is utilized to assure the pressure suppression function of the containment is maintained while the reactor is at pressure_

(For additional information about the Pressure Suppression Pressure see the EOP User's Guide.)

1 00 1-37.8 Rev. 4 Page 30 of 58/

  • STEPS PC/P-14 through PC/P-16 (continued)

A note is added to remind the operator that rapid depressurization per the reactor pressure control guidance of the Reactor Vessel Control Procedure may be allowed prior to the direction to Emergency Depressurize. See discussion on Step SPIT-11 .

  • 1001-37.8 Rev. 4 Page 31 of 581
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