ML13354B924

2013, Dresden Initial License Exam, Written Exam References

ML13354B924
Person / Time
Site:	Dresden
Issue date:	11/25/2013
From:	Walton R NRC/RGN-III/DRS/OLB
To:
Walton R
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ML11355A018	List: ML13347B096 ML13354B850 ML13354B863 ML13354B869 ML13354B870 ML13354B924 ML13354C049
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Download: ML13354B924 (60)

Text

SLC System 3.1.7 Figure 3.1.7-1 (page 1 of 1)

Sodium Pentaborate Volume Requirements Dresden 2 and 3 3.1.7-5 Amendment No. 237/230

SLC System 3.1.7 150 (14%, 150 F) (16.5%, 150 F) 140 130 120 Acceptable Operating Region Temperature (F) 110 100 90 (16.5%, 83 F) 80 (14%, 73.5 F) 70 60 13 14 15 16 17 Sodium Pentaborate Concentration (% by Weight)

Figure 3.1.7-2 (page 1 of 1)

Sodium Pentaborate Temperature Requirements Dresden 2 and 3 3.1.7-6 Amendment No. 237/230

RADIOLOGICAL SURVEY FORM Training1.ppt SURVEY NUMBER RW P NUMBER 13-0029 13-0010 BUILDING ELEVATION AREA/ROOM/SYSTEM DATE TIME Training 603 Rad Worker Exercise Area 10-13-13 8:00 PURPOSE  % POWER Weekly 100 Legend:

All radiation readings are in mrem/hr and all smears are in dpm/100cm2 unless otherwise noted.

O - Smear - Alpha - Beta - Gamma - neutron */ - Contact/30cm [ ] - Hot Spot XXXX - Boundary Bkgd - Background SOP - Step Off Pad MDA- Minimum Detectable Activity CA - Contamination Area HPZ - Hot Particle Zone RA - Radiation Area HRA - High Radiation Area LHRA - Locked High Radiation Area ARA - Airborne Radiation Area Smear Smear dpm/100cm2

1. Location 1 SOP <32 <30 2 Floor 9,975 <30 3 Floor 19,476 <30 4 Floor 18,958 <30 5 Floor 120,745 <30 0.5 0.5 0.5 6 Floor 81,047 <30 10 7 Floor 1,788 <30 SOP SOP 8 Floor 9,267 <30 XXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXXX 9 Floor 9,546 <30 RA/CA 10 Floor <32 <30 MDA 32 30 9 8 20 7 20 6

MU46 XXXXXXXXXXXXXXXXXX 45 XXXXXXXXXXXXXXXXXXXXXXXXXX MU-16 60

• 2000/120 90 30 90 HRA/HCA MU26 3

5 4

Pump 30 XXXXXXXXXXXXXXXXXX Cooler 2

1 INSTRUMENTS USED PREPARED BY:

MODEL NUMBER NAME (Print) SIGNATURE DATE LI NUMBER CAL DUE DATE RSO-5 2.7.148 11/10/13 APPROVED BY:

LB-5100 2.12.26 10/31/13 NAME (Print) SIGNATURE DATE N REVIEW ED BY:

NAME (Print) SIGNATURE DATE A

PAGE_ OF_ PAGES

CATEGORY 1 UNIT 2(3)

DOA 0202-01 REVISION 38 D: Subsequent Operator Actions (NOMINAL FEEDWATER HEATING) (HARD CARD):

D START At least one OPRM Is OPRM TS 3.3.1.3 channel operable per NO Required Action A.3 RPS trip system? or B.1 in effect?

YES NO YES Is running Pp speed Scram reactor 68% AND YES and enter FCL 67.8%? DGP 02-03 NO Closed tripped recirc Monitor MSL &

Is FCL YES pump discharge valve(s) off gas rad

< 55%? MO 2(3)-0202-5A monitors for MO 2(3)-0202-5B increased activity.

NO AFTER 5 minutes, IF SDC NOT in IF running recirc operation, THEN pump speed is > 77% Notify QNE to open recirc pump AND FCL 90.0%, monitor core discharge valve THEN reduce recirc parameters.

previously closed.

pump speed to 77%

(75 to 79%)

CAUTION Insert CRAM Rods DO NOT reduce Rx per DGP 03-04 to power to < 10% Notify Chemistry reduce Rx power with Recirc Flow to take samples to 25 to 30%

per TS and ODCM if power change was 20% thermal in Reduce running Recirc one hour.

Pp speed to:

Unit 2 < 35%

Unit 3 < 60%

Continue in Subsequent Operator Actions (Continued) on page 5.

3 of 12

CATEGORY 1 UNIT 2(3)

DOA 0202-01 REVISION 38 D: Subsequent Operator Actions (REDUCED FEEDWATER HEATING) (HARD CARD):

START At least one OPRM Is OPRM TS 3.3.1.3 NO Required Action A.3 channel operable per RPS trip system? or B.1 in effect?

YES NO YES Is running Pp speed 91%? Scram reactor YES AND FCL 65.3%? and enter DGP 02-03 NO Closed tripped recirc Monitor MSL &

Is FCL YES pump discharge valve(s) off gas rad

< 55%? MO 2(3)-0202-5A monitors for MO 2(3)-0202-5B increased NO activity.

IF running recirc AFTER 5 minutes, IF pump speed is > 77% SDC NOT in AND FCL 90.0%, operation, THEN Notify QNE to THEN reduce recirc open recirc pump monitor core pump speed to 77% discharge valve parameters.

(75 to 79%) previously closed.

CAUTION Insert CRAM Rods Do NOT reduce Rx per DGP 03-04 to power to < 10%

with Recirc Flow Notify Chemistry reduce Rx power to 25 to 30% to take samples per TS and ODCM if power change was 20% thermal in Reduce running recirc one hour.

pump speed to:

Unit 2 < 35%

Unit 3 < 60%

Continue in Subsequent Operator Actions (Continued) on page 5.

4 of 12

LS-AA-1120 Revision 15 Page 13 of 103 REPORTABLE EVENT RAD 1.4:

Liquid Effluent Release Requirement: 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 20.2203 (a)(3)

§ 50.73(a)(2)(viii)(B): The licensee shall report ... any liquid effluent release that, when averaged over a time period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, exceeds 20 times the applicable concentrations specified in Appendix B to 10 CFR 20, Table 2, Column 2, at the point of entry into the receiving waters (i.e., unrestricted area) for all radionuclides except tritium and dissolved noble gases.

§ 20.2203(a)(3): ... each licensee shall submit a written report ... after learning of ...

levels of radiation or concentrations of radioactive material in -

(i) A restricted area in excess of any applicable limit in the license; or (ii) An unrestricted area in excess of 10 times any applicable limit set forth in this part or in the license (whether or not involving exposure of any individual in excess of the limits in 10 CFR 20.1301).

Time Required Notification(s):

Limit NONE No immediate notification is required for this Reportable Event.

Time Required Written Report(s):

Limit 30 DAYS Submit a written report to the NRC within 30 days of discovery of levels of radiation or concentrations of radioactive material in excess of the limits of

§ 20.2203(a)(3). Prepare and submit the report in accordance with the requirements of § 20.2203(b) and (c). [10 CFR 20.2203(a)(3)]

60 DAYS Submit a Licensee Event Report to the NRC within 60 days of discovery of the occurrence of any liquid effluent release that exceeded the limits of

§ 50.73(a)(2)(viii)(B), if the release occurred within 3 years of the date of discovery. [10 CFR 50.73(a)(1), 10 CFR 50.73(a)(2)(viii)(B)]

Reportability Reference Manual

LS-AA-1120 Revision 15 Page 14 of 103 REPORTABLE EVENT RAD 1.4 (Contd)

Discussion:

o NRC guidance on this Reportable Event is provided in NUREG 1022, Revision 2, Section 3.2.9.

o The occurrence of this event may require activation of the Emergency Plan. In that case, notification will be made per the Emergency Plan, and a duplicate notification per this Reportable Event is not required. [See SAF 1.1]

o "Unrestricted Area" means any area at or beyond the site boundary, access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, and any area within the site boundary used for residential quarters or industrial, commercial, institutional and recreational facilities.

o The location used as the point of release for calculation purposes should be determined using the expanded definition, listed above, for an unrestricted area as specified in NUREG 0133 to maintain consistency with the TS.

Related Reportable Events:

o RAD 1.1, Events Involving Byproduct, Source or Special Nuclear Material That Cause or Threaten to Cause Significant Exposure or Release o RAD 1.2, Events Involving Loss of Control of Licensed Material That Cause or Threaten to Cause Exposure or Release

References:

o NUREG 1022, Revision 2 o NUREG-0133, Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants o NRC Generic Letter 85-19, Reporting Requirements on Primary Coolant Iodine Spikes, September 1985 o 10 CFR 50.73 o 10 CFR 20.2203 Reportability Reference Manual

RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.1.1-1.

ACTIONS

---

NOTES -----------------------------------

1. Separate Condition entry is allowed for each channel.

2. When Functions 2.b and 2.c channels are inoperable due to APRM indication not within limits, entry into associated Conditions and Required Actions may be delayed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> if the APRM is indicating a lower power value than the calculated power, and for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if the APRM is indicating a higher power value than the calculated power.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> channels inoperable. trip.

OR A.2 Place associated trip 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> system in trip.

B. One or more Functions B.1 Place channel in one 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with one or more trip system in trip.

required channels inoperable in both OR trip systems.

B.2 Place one trip system 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> in trip.

(continued)

Dresden 2 and 3 3.3.1.1-1 Amendment No. 185/180

RPS Instrumentation 3.3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. One or more Functions C.1 Restore RPS trip 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with RPS trip capability.

capability not maintained.

D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, Table 3.3.1.1-1 for B, or C not met. the channel.

E. As required by E.1 Reduce THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Required Action D.1 to < 38.5% RTP and referenced in Table 3.3.1.1-1.

F. As required by F.1 Be in MODE 2. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Required Action D.1 and referenced in Table 3.3.1.1-1.

(continued)

Dresden 2 and 3 3.3.1.1-2 Amendment No. 239/232

RPS Instrumentation 3.3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME G. As required by G.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Required Action D.1 and referenced in Table 3.3.1.1-1.

H. As required by H.1 Initiate action to Immediately Required Action D.1 fully insert all and referenced in insertable control Table 3.3.1.1-1. rods in core cells containing one or more fuel assemblies.

Dresden 2 and 3 3.3.1.1-3 Amendment No. 185/180

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS

---

NOTES -----------------------------------

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.2 ------------------NOTE-------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER 25% RTP.

Verify the absolute difference between In accordance the average power range monitor (APRM) with the channels and the calculated power is Surveillance 2% RTP. Frequency Control Program SR 3.3.1.1.3 Adjust the channel to conform to a In accordance calibrated flow signal. with the Surveillance Frequency Control Program (continued)

Dresden 2 and 3 3.3.1.1-4 Amendment No. 237/230

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.4 ------------------NOTE-------------------

Not required to be performed when entering MODE 2 from MODE 1 until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after entering MODE 2.

Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.5 Perform a functional test of each RPS In accordance automatic scram contactor. with the Surveillance Frequency Control Program SR 3.3.1.1.6 Verify the source range monitor (SRM) and Prior to fully intermediate range monitor (IRM) channels withdrawing overlap. SRMs SR 3.3.1.1.7 ------------------NOTE-------------------

Only required to be met during entry into MODE 2 from MODE 1.

Verify the IRM and APRM channels overlap. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.8 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program (continued)

Dresden 2 and 3 3.3.1.1-5 Amendment No. 237/230

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.9 Calibrate the local power range monitors. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.10 Deleted.

SR 3.3.1.1.11 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.12 Calibrate the trip units. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.13 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.14 Verify Turbine Stop ValveClosure and In accordance Turbine Control Valve Fast Closure, Trip with the Oil PressureLow Functions are not Surveillance bypassed when THERMAL POWER is Frequency 38.5% RTP. Control Program (continued)

Dresden 2 and 3 3.3.1.1-6 Amendment No. 237/230

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.15 ------------------NOTES------------------

1. Neutron detectors are excluded.

2. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after entering MODE 2.

3. For Function 2.b, not required for the flow portion of the channels.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.16 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.17 ------------------NOTES------------------

1. Neutron detectors are excluded.

2. For Function 1.a, not required to be performed when entering MODE 2 from MODE 1 until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after entering MODE 2.

Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program (continued)

Dresden 2 and 3 3.3.1.1-7 Amendment No. 237/230

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.1.18 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.1.1.19 ------------------NOTES------------------

Neutron detectors are excluded.

Verify the RPS RESPONSE TIME is within In accordance limits. with the Surveillance Frequency Control Program Dresden 2 and 3 3.3.1.1-8 Amendment No. 237/230

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

1. Intermediate Range Monitors

a. Neutron FluxHigh 2 3 G SR 3.3.1.1.1 121/125 SR 3.3.1.1.4 divisions of SR 3.3.1.1.5 full scale SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.17 SR 3.3.1.1.18 5(a) 3 H SR 3.3.1.1.1 121/125 SR 3.3.1.1.4 divisions of SR 3.3.1.1.5 full scale SR 3.3.1.1.17 SR 3.3.1.1.18

b. Inop 2 3 G SR 3.3.1.1.4 NA SR 3.3.1.1.5 SR 3.3.1.1.18 5(a) 3 H SR 3.3.1.1.4 NA SR 3.3.1.1.5 SR 3.3.1.1.18

2. Average Power Range Monitors

a. Neutron FluxHigh, 2 2 G SR 3.3.1.1.1 17.1% RTP Setdown SR 3.3.1.1.4 SR 3.3.1.1.5 SR 3.3.1.1.7 SR 3.3.1.1.9 SR 3.3.1.1.15 SR 3.3.1.1.18

b. Flow Biased Neutron 1 2 F SR 3.3.1.1.1 0.56 W FluxHigh SR 3.3.1.1.2 67.4% RTP and SR 3.3.1.1.3 122% RTP(b)

SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.11 SR 3.3.1.1.15 SR 3.3.1.1.17 SR 3.3.1.1.18 SR 3.3.1.1.19 (continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) 0.56 W + 63.2% and 118.5% RTP when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating."

Dresden 2 and 3 3.3.1.1-9 Amendment No. 237/230

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

2. Average Power Range Monitors (continued)

c. Fixed Neutron 1 2 F SR 3.3.1.1.1 122% RTP Flux-High SR 3.3.1.1.2 SR 3.3.1.1.5 SR 3.3.1.1.9 SR 3.3.1.1.11 SR 3.3.1.1.15 SR 3.3.1.1.18 SR 3.3.1.1.19

d. Inop 1,2 2 G SR 3.3.1.1.5 NA SR 3.3.1.1.9 SR 3.3.1.1.11 SR 3.3.1.1.18

3. Reactor Vessel Steam 1,2 2 G SR 3.3.1.1.1 1045 psig Dome PressureHigh SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.12 SR 3.3.1.1.17 SR 3.3.1.1.18 SR 3.3.1.1.19

4. Reactor Vessel Water 1,2 2 G SR 3.3.1.1.1 2.65 inches LevelLow SR 3.3.1.1.5 SR 3.3.1.1.11 SR 3.3.1.1.12 SR 3.3.1.1.17 SR 3.3.1.1.18 SR 3.3.1.1.19

5. Main Steam Isolation 1 8 F SR 3.3.1.1.5 9.5% closed ValveClosure SR 3.3.1.1.11 SR 3.3.1.1.17 SR 3.3.1.1.18 SR 3.3.1.1.19

6. Drywell PressureHigh 1,2 2 G SR 3.3.1.1.5 1.94 psig SR 3.3.1.1.11 SR 3.3.1.1.13 SR 3.3.1.1.18 SR 3.3.1.1.19 (continued)

Dresden 2 and 3 3.3.1.1-10 Amendment No. 239/232

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 3 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

7. Scram Discharge Volume Water LevelHigh

a. Thermal Switch 1,2 2 G SR 3.3.1.1.1(c) 37.9 gallons (Unit 2) SR 3.3.1.1.5 (Unit 2)

Level Indicating SR 3.3.1.1.11 38.7 gallons Switch (Unit 3) SR 3.3.1.1.12(c) (Unit 3)

SR 3.3.1.1.17 SR 3.3.1.1.18 5(a) 2 H SR 3.3.1.1.1(c) 37.9 gallons SR 3.3.1.1.5 (Unit 2)

SR 3.3.1.1.11 38.7 gallons SR 3.3.1.1.12(c) (Unit 3)

SR 3.3.1.1.17 SR 3.3.1.1.18

b. Differential 1,2 2 G SR 3.3.1.1.5 37.9 gallons Pressure Switch SR 3.3.1.1.11 (Unit 2)

SR 3.3.1.1.12 38.7 gallons SR 3.3.1.1.17 (Unit 3)

SR 3.3.1.1.18 5(a) 2 H SR 3.3.1.1.5 37.9 gallons SR 3.3.1.1.11 (Unit 2)

SR 3.3.1.1.12 38.7 gallons SR 3.3.1.1.17 (Unit 3)

SR 3.3.1.1.18

8. Turbine Stop 38.5% RTP 4 E SR 3.3.1.1.5 9.5% closed Valve-Closure SR 3.3.1.1.11 SR 3.3.1.1.14 SR 3.3.1.1.17 SR 3.3.1.1.18 SR 3.3.1.1.19

9. Turbine Control Valve 38.5% RTP 2 E SR 3.3.1.1.5 466 psig Fast Closure, Trip Oil SR 3.3.1.1.11 PressureLow SR 3.3.1.1.14 SR 3.3.1.1.17 SR 3.3.1.1.18 SR 3.3.1.1.19

10. Turbine Condenser 1 2 F SR 3.3.1.1.5 20.5 inches VacuumLow SR 3.3.1.1.11 Hg vacuum SR 3.3.1.1.13 SR 3.3.1.1.18 SR 3.3.1.1.19

11. Reactor Mode Switch 1,2 1 G SR 3.3.1.1.16 NA Shutdown Position SR 3.3.1.1.18 5(a) 1 H SR 3.3.1.1.16 NA SR 3.3.1.1.18

12. Manual Scram 1,2 1 G SR 3.3.1.1.8 NA SR 3.3.1.1.18 5(a) 1 H SR 3.3.1.1.8 NA SR 3.3.1.1.18 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(c) Specified SR performance only required for Unit 3.

Dresden 2 and 3 3.3.1.1-11 Amendment No. 239/232

TRM Fire Water Supply System 3.7.i 3.7 PLANT SYSTEMS 3.7.i Fire Water Supply System TLCO 3.7.i The Fire Water Supply System shall be OPERABLE with:

1. A flow path for the Unit 2/3 fire pump capable of taking suction from the Unit 2/3 intake canal and aligned to discharge to the fire water supply header;

2. A flow path to the Unit 1 fire pump capable of taking suction from the Unit 1 intake canal and aligned to discharge to the fire water supply header;

3. Automatic initiation logic for each fire pump;

4. Fire water supply header piping with sectional control valves to the yard loop, the front valve ahead of the water flow alarm device on each sprinkler or water spray system, and the standpipe system.

APPLICABILITY: At all times.

ACTIONS

---

NOTE------------------------------------------

Separate Condition entry is allowed for each fire pump.

CONDITION REQUIRED ACTION COMPLETION TIME A. One fire pump or water A.1 Restore equipment to 7 days supply inoperable. OPERABLE status.

OR A.2 Prepare a corrective 7 days action program report.

Dresden 2 and 3 3.7.i-1 Revision 0

TRM Fire Water Supply System 3.7.i ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Two fire pumps or water B.1 Establish a backup 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> supplies inoperable. water supply.

C. Required Action B.1 and C.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met. AND C.2 Be in MODE 4. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.7.i.1 Verify the electrolyte level of each battery 31 days for each diesel driven fire pump is above the plates.

TSR 3.7.i.2 Verify the overall battery voltage for each 31 days diesel driven fire pump is > 24 volts.

TSR 3.7.i.3 Verify the unit 1 diesel driven fire pump 31 days fuel storage day tank contains > 150 gallons of fuel and the unit 2/3 diesel driven fire pump fuel storage day tank contains > 208 gallons of fuel.

(continued)

Dresden 2 and 3 3.7.i-2 Revision 46

TRM Fire Water Supply System 3.7.i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.7.i.4 Start each fire pump from ambient conditions and 31 days operate each fire pump on recirculation flow for

> 30 minutes.

TSR 3.7.i.5 Verify a sample of fuel from the diesel driven 92 days fire pump fuel storage tank, obtained in accordance with ASTM-D4057-95, is within the acceptable limits specified in Table 1 of ASTM-D975-98b with respect to viscosity, water content, and sediment.

TSR 3.7.i.6 --------------------NOTE------------------------

Not applicable for nickel cadmium batteries.

Verify the specific gravity of each battery for 92 days the diesel driven fire pump is appropriate for continued service of the battery.

TSR 3.7.i.7 Verify that each valve in the flow path that is 184 days not locked, sealed, or otherwise secured in position is in the correct position.

TSR 3.7.i.8 Perform a system flush. 12 months TSR 3.7.i.9 Verify the battery and battery racks for each 18 months diesel driven fire pump show no visual indication of physical damage or abnormal deterioration.

(continued)

Dresden 2 and 3 3.7.i-3 Revision 0

TRM Fire Water Supply System 3.7.i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.7.i.10 Verify the battery-to-battery and terminal 18 months connections for each diesel driven fire pump are clean, tight, free of corrosion and coated with anti-corrosion material.

TSR 3.7.i.11 Perform a system functional test, which includes 18 months simulated automatic actuation of the system throughout its operating sequence and:

a. Verify that each automatic valve in the flow path actuates to its correct position;

b. Verify that the Unit 2/3 fire pump develops

> 3000 gpm at a system pressure > 126 psig; and

c. Verify that the Unit 1 fire pump develops

> 2500 gpm at a system pressure > 136 psig.

TSR 3.7.i.12 Perform a system functional test in accordance 18 months with NFPA 20-1976.

TSR 3.7.i.13 Cycle each testable valve in the flow path 18 months through one complete cycle.

(continued)

Dresden 2 and 3 3.7.i-4 Revision 52

TRM Fire Water Supply System 3.7.i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.7.i.14 Perform a flow test of the system in accordance 36 months with the "Tests of Water Supplies" chapter of the Fire Protection Handbook published by the National Fire Protection Association.

TSR 3.7.i.15 Inspect the diesel of each diesel driven fire 72 months pump in accordance with procedures prepared in conjunction with the manufacturer recommendations for the class of service.

Dresden 2 and 3 3.7.i-5 Revision 0

TRM Water Suppression Systems 3.7.j 3.7 PLANT SYSTEMS 3.7.j Water Suppression Systems TLCO 3.7.j The Water Suppression Systems shown in Table T3.7.j-1 shall be OPERABLE.

APPLICABILITY: Whenever equipment protected by the suppression systems is required to be OPERABLE.

ACTIONS

---

NOTE------------------------------------------

Separate Condition entry is allowed for each Water Suppression System.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Water A.1 -------NOTE---------

Suppression Systems 1. Not applicable for inoperable. Unit 2/3 mezzanine 534 ft elevation area and hydrogen seal oil areas.

2. Not applicable for inaccessible areas.

3. Not applicable for areas with OPERABLE detection.

Establish an hourly 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire watch patrol with backup fire suppression equipment.

AND (continued)

Dresden 2 and 3 3.7.j-1 Revision 0

TRM Water Suppression Systems 3.7.j ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 -------NOTE----------

1. Not applicable for Unit 2/3 mezzanine 534 ft elevation area and hydrogen seal oil areas.

2. Not applicable for inaccessible areas.

3. Applicable for areas with OPERABLE detection.

Establish a once per 8 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> hour fire watch patrol with backup fire suppression equipment.

AND A.3 -------NOTE----------

1. Not applicable for Unit 2/3 mezzanine 534 ft elevation area and hydrogen seal oil areas.

2. Not applicable for accessible areas.

Establish a once per 8 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> hour fire watch patrol with backup fire suppression equipment.

AND Dresden 2 and 3 3.7.j-2 Revision 70

TRM Water Suppression Systems 3.7.j CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.4 -------NOTE----------

Applicable for Unit 2/3 mezzanine 534 ft elevation area and hydrogen seal oil areas.

Establish a continuous 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire watch with backup fire suppression equipment.

AND A.5.1 Restore the system to 14 days OPERABLE status.

OR A.5.2 Prepare a corrective 14 days action program report.

OR A.6 -------NOTE-------

Only Applicable following completion and implementation of approved technical evaluation.

Enter 3.0.g Immediately Dresden 2 and 3 3.7.j-3 Revision 70

TRM Water Suppression Systems 3.7.j SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.7.j.1 Verify each manual, power operated, or automatic 184 days valve in the flow path is in the correct position.

TSR 3.7.j.2 Cycle each testable valve in the flow path 12 months through one complete cycle of full travel.

TSR 3.7.j.3 Perform a system functional test, which includes 18 months simulated automatic actuation of the system, and verify that automatic valves in the flow path actuate to their correct positions.

(continued)

Dresden 2 and 3 3.7.j-4 Revision 0

TRM Water Suppression Systems 3.7.j SURVEILLANCE FREQUENCY TSR 3.7.j.4 ----------------NOTES----------------------- 18 months Not required to be performed for sprinkler piping inaccessible during plant operations.

Perform visual inspection of the sprinkler piping to verify its integrity.

TSR 3.7.j.5 ----------------NOTES----------------------- 18 months Not required to be performed for nozzles inaccessible during plant operations.

Perform visual inspection of each nozzles spray area to verify that the spray pattern is not obstructed.

TSR 3.7.j.6 ----------------NOTES-----------------------

Only required to be performed for sprinkler piping inaccessible during plant operations.

Perform visual inspection of the sprinkler 24 months piping to verify its integrity.

TSR 3.7.j.7 ----------------NOTES-----------------------

Only required to be performed for nozzles inaccessible during plant operations.

Perform visual inspection of each nozzles spray 24 months area to verify that the spray pattern is not obstructed.

(continued)

Dresden 2 and 3 3.7.j-5 Revision 0

TRM Water Suppression Systems 3.7.j SURVEILLANCE FREQUENCY TSR 3.7.j.8 Perform a flow test through each open head 36 months spray nozzle to verify the discharge pattern of the nozzles and that each open head spray nozzle is properly aimed.

Dresden 2 and 3 3.7.j-6 Revision 57

TRM Water Suppression Systems 3.7.j Table T3.7.j-1 (page 1 of 2)

Water Suppression Systems I. Preaction sprinkler system

1. Unit 2 HPCI Room Preaction System.

2. Unit 3 HPCI Room Preaction System.

3. Instrument Air Compressor 2-4706 Preaction System.

4. Unit 2 Hatchways and Stairways El. 570' - 0" and 589' - 0" Cols 43, 44 42 N to M Preaction System.

5. Unit 3 Around Hatchways el. 570' - 0" and 589' - 0" Cols 45, 46 - N M Preaction System.

6. Mezz. Floor Cable Concentration Area Col. Row 33 to 38 G to H Preaction In-Tray Sprinkler System.

II. Wet pipe sprinkler systems

1. Unit 2/3 Diesel Generator Day Tank Wet Pipe Sprinkler System

2. Unit 2 ACAD Air Compressor Wet Pipe Sprinkler System

3. Unit 3 ACAD Air Compressor Wet Pipe Sprinkler System

4. Unit 2 Condensate Pump Room Wet Pipe Sprinkler System *

5. Unit 3 Condensate Pump Room Wet Pipe Sprinkler System *

6. Unit 2 CRD and CCSW Pumps Wet Pipe Sprinkler System *

7. Unit 3 CRD and CCSW Pumps Wet Pipe Sprinkler System *

8. Clean and Dirty Oil Tank Room Wet Pipe Sprinkler System *

9. Unit 2 Below Turbine South Side Wet Pipe Sprinkler System *

10. Unit 3 Below Turbine South Side Wet Pipe Sprinkler System *

11. Unit 2 Reactor Feed Pump and Speed Increaser Wet Pipe Sprinkler System *

12. Unit 3 Reactor Feed Pump and Speed Increaser Wet Pipe Sprinkler System *

13. Unit 2 Below Turbine North Side Wet Pipe Sprinkler System *

14. Unit 3 Below Turbine North Side Wet Pipe Sprinkler System *

15. Unit 2 Trackway and area between Column 33 and 35 F and G Wet Pipe Sprinkler System *

16. Unit 3 Trackway Wet Pipe Sprinkler System *

17. Unit 2 Diesel Generator Day Tank Wet Pipe Sprinkler System *

18. Unit 3 Diesel Generator Day Tank Wet Pipe Sprinkler System *

19. Unit 2 and 3 Cable Tunnel Wet Pipe Sprinkler System

20. Unit 2 and 3 Turbine Building Common Mezzanine Area Wet Pipe Sprinkler System *

21. Unit 2 Wet Pipe Sprinkler System at Ceiling above Stator Cooling and H2 Seal Oil *

22. Unit 3 Wet Pipe Sprinkler System at Ceiling above Stator Cooling and H2 Seal Oil *

23. Unit 2 MG Set 2A-251 and 2B-251 Fluid Drive Wet Pipe Sprinkler System *

24. Unit 3 MG Set 3A-251 and 3B-251 Fluid Drive Wet Pipe Sprinkler System *

25. Unit 2 Turbine Bearing Lift Pump (Col-36) Wet Pipe Sprinkler System *

26. Unit 2 Turbine Bearing Lift Pump (Col-41) Wet Pipe Sprinkler System *

27. Unit 3 Turbine Bearing Lift Pump (Col-47) Wet Pipe Sprinkler System *

28. Unit 3 Turbine Bearing Lift Pump (Col-53) Wet Pipe Sprinkler System *

29. Unit 2 and 3 Turbine Building Common Area Corridor Elev. 517-6 Wet Pipe Sprinkler System

30. Unit 2/3 Cribhouse Lower Level Wet Pipe Sprinkler System

31. Unit 2/3 Cribhouse Upper East Wet Pipe Sprinkler System *

32. Unit 2/3 Cribhouse Upper West Wet Pipe Sprinkler System *

33. Unit 2 Mechanical Opening Outside of Safe Shutdown Heat Exchanger Room Wet Pipe Sprinkler System

34. Unit 3 Mechanical Opening Outside of Safe Shutdown Heat Exchanger Room Wet Pipe Sprinkler System

35. Unit 2 Ladderway 587-0, Col. 38 M, Wet Pipe Sprinkler System

36. Unit 3 Ladderway 587-0, Col. 50 M, Wet Pipe Sprinkler System

37. Unit 1 Diesel Fire Pump Wet Pipe Sprinkler System *

38. Unit 2 Oil Storage Area Wet Pipe Sprinkler System *

39. Unit 1 West Aux Bay South Wet Pipe Sprinkler System *

40. Unit 1 West Aux Bay North Wet Pipe Sprinkler System *

41. E.H.C. Units 2-5614 and 3-5614 Wet Pipe Sprinkler System (continued)

• These Wet Pipe Sprinkler Systems do not have fire detection systems in the same area.

Dresden 2 and 3 3.7.j-7 Revision 60

TRM Water Suppression Systems 3.7.j Table T3.7.j-1 (page 2 of 2)

Water Suppression Systems III. Water spray (open head) systems

1. Unit 2 Turbine Oil Reservoir Deluge System.

2. Unit 3 Turbine Oil Reservoir Deluge System.

3. Unit 2 Hydrogen Seal Unit Deluge System.

4. Unit 3 Hydrogen Seal Unit Deluge System.

5. Diesel Fire Pump and Day Tank Deluge System.

6. Main Power Transformer #2 Deluge System.

7. Main Power Transformer #3 Deluge System.

8. Auxiliary Transformer #21 Deluge System.

9. Auxiliary Transformer #31 Deluge System.

10. Reserve Auxiliary Transformer #22 Deluge System.

11. Reserve Auxiliary Transformer #32 Deluge System.

12. Unit 2 Bus Duct Penetration Deluge System.

13. Unit 3 Bus Duct Penetration Deluge System.

14. Cribhouse Cable Tray Open Head Water Spray System.

15. 2/3 D.G. Cooling Water Pump Deluge System.

Dresden 2 and 3 3.7.j-8 Revision 52

RP-AA-203 Revision 3 Page 1 of 12 Level 3 - Information Use EXPOSURE CONTROL AND AUTHORIZATION

1. PURPOSE 1.1. This procedure describes the program by which the Radiation Protection Department evaluates and controls personnel occupational exposure.

2. TERMS AND DEFINITIONS 2.1. Absent/ No record (A): Dose type used for monitoring periods entered under 10CFR20 for which hard copy dose documentation of dose data from the monitoring licensee is unobtainable. The quantity associated with A dose type is always XX or intentionally left blank.

2.2. Administrative Dose Control Level (ADCL): An Exelon established dose guideline established to prevent personnel form exceeding the Federal dose limits and to help ensure equitable distribution of dose among workers with similar jobs.

2.3. Emergency Exposure: Exposure received during lifesaving, protection of valuable property, protection of large populations, or any immediate action taken in response to a situation or occurrence of a serious nature developing suddenly and unexpectedly.

2.4. High Lifetime Exposure: The cumulative TEDE (routine plus any PSE) in rem that is equal to or exceeds an individuals age in years.

2.5. Lens Dose Equivalent (LDE): The external exposure of the lens of the eye and is taken at the dose equivalent at a tissue depth of 300 mg/cm2.

2.6. Planned Special Exposure (PSE): An infrequent exposure to radiation, separate from and in addition to the annual exposure limits (Refer to Table 1).

2.7. Shallow Dose Equivalent (SDE): The external exposure of the skin or an extremity taken at a tissue depth of 7 mg/cm2.

2.8. Total Effective Dose Equivalent (TEDE): The sum of the deep dose equivalent (external exposure) and the committed effective dose equivalent (internal exposure).

2.9. Total Organ Dose Equivalent (TODE): The sum of the deep dose equivalent and the committed dose equivalent to the organ receiving the highest dose.

3. RESPONSIBILITIES

RP-AA-203 Revision 3 Page 2 of 12 3.1. Responsibility for approval of exposures in excess of administrative dose control levels resides with the Radiation Protection Manager (RPM), the Station/Plant Manager, and the Site Vice President.

4. MAIN BODY 4.1. Limitations 4.1.1. Exposures shall not exceed the 10CFR20 Exposure Limits as described in Table 1, NRC Exposure Limits.

TABLE 1 - NRC EXPOSURE LIMITS Individual and Limit Type TEDE LDE SDE TODE (1) Occupational Worker, Minor, Routine Annual 0.5 rem 1.5 rem 5 rem 5 rem (2) Occupational Worker, Adult, Routine Annual 5 rem 15 rem 50 rem 50 rem (3) Occupational Worker, 5 rem 15 rem 50 rem 50 rem Adult, PSE Annual (4) Occupational Worker, 25 rem 75 rem 250 rem 250 rem Adult, PSE Lifetime Dose equivalent to the embryo/fetus of 500 mrem, or 50 additional (5) Declared Pregnant mrem if the dose equivalent to the embryo/fetus exceeds 450 mrem at Woman the time the female declares, in writing, her pregnancy. (Note: The dose equivalent to the embryo/fetus equals the DDE to the declared pregnant woman plus the dose equivalent to the embryo/fetus from radionuclides in the embryo/fetus plus the dose equivalent to the embryo/fetus from radionuclides in the declared pregnant woman.)

(6) Member of The Public The total effective dose equivalent to individual members of the public shall not exceed 100 mrem in a year.

RP-AA-203 Revision 3 Page 3 of 12 NOTE: Any request to raise the administrative dose control level for a minor shall be approved and documented by the Radiation Protection Manager.

4.1.2. Administrative dose control levels have been established for Total Effective Dose Equivalent Limits as follows:

2000 mrem routine cumulative TEDE/yr.

200 mrem TEDE for minors.

NOTE: In the Midwest Regional Operating Group, controls for High Lifetime Exposure are not applicable for non-Exelon employees due to the limitations of the Exposure Tracking System.

4.1.3. An administrative dose control level of 1000 mrem TEDE plus PSE has been established for employees with High Lifetime Exposure.

4.1.4. The Radiation Protection Manger shall review an individuals occupational exposure when the dose equivalent reaches 80% of the NRC Limits for Lens Dose Equivalent (LDE), Shallow Dose Equivalent (SDE), and Total Organ Dose Equivalent (TODE).

The 80% threshold values are as follows:

12 rem LDE.

40 rem SDE.

40 rem TODE.

4.1.5. If an individual's current year dose history documentation includes an absent/ no record (A) dose type, then REDUCE the individuals allowable exposure (normally 2000 mrem TEDE for the year) by 1250 mrem TEDE for each quarter of the current year for which dose history documentation is absent/ no record (A), until all of that dose is resolved.

4.1.6. If an individual is suspected of exceeding any of the NRC exposure limits in Table 1, then PROHIBIT the individual from entering the RCA until a detailed evaluation of the individual's actual dose equivalent has been conducted. Future access will depend upon the results of the evaluation.

1. If an exposure in excess of the applicable exposure limit has occurred, then PROHIBIT the individual from entering the RCA until the end of the current calendar year.

2. If an exposure in excess of the applicable exposure limit has not occurred, then the individual may be permitted to re-enter the RCA.

RP-AA-203 Revision 3 Page 4 of 12 4.1.7. During a condition where the Generating Stations Emergency Plan has been initiated, emergency exposure authorizations shall be performed in accordance with the stations Emergency Plan Implementing Procedures.

4.2. Authorization To Raise Administrative Dose Control Levels (ADCLs) 4.2.1. USE Attachment 1, Dose Control Level Extension Form, or a computerized equivalent, to authorize exposures for adult individuals in excess of 2000 mrem routine TEDE in a year.

4.2.2. A supervisor from the department requesting approval shall complete Section I of Attachment 1 and submit the request to the Radiation Protection Department indicating:

The name, identification number, and signature of the individual for whom a dose extension is being requested.

Whether or not other qualified individuals with lower dose are available to perform the work.

A detailed explanation of why the dose extension is necessary.

The requested annual TEDE limit for the individual (expressed in 500 mrem increments, i.e. 2500 mrem, 3000 mrem, etc.)

4.2.3. The Radiation Protection Department shall complete Section II Attachment 1 or computerized equivalent.

4.2.4. Pending investigations or calculations of internal exposure shall be reviewed and evaluated to determine the individuals TEDE.

4.2.5. Non-Exelon Nuclear and non-ROG dose equivalent shall be included in the dose to determine the workers current TEDE.

NOTE: An individual shall not be approved to receive greater than 2000 mrem TEDE if that person has any absent/ no record (A) dose equivalent for the year.

4.2.6. To raise the ADCL up to and including 3000 mrem TEDE in a calendar year, written approval is required by the Radiation Protection Manager and the work group supervisor.

4.2.7. To raise the ADCL to between 3001 and 4000 mrem TEDE in a calendar year, written approval is required by the Radiation Protection Manger, a work group supervisor, and the Station/Plant Manager.

NOTE: An individual being considered for approval greater than 4000 mrem TEDE for the year should not have significant estimated dose equivalent.

4.2.8. To raise the ADCL above 4000 mrem, not to exceed 5000 mrem, written approval is required by the Site Vice President.

RP-AA-203 Revision 3 Page 5 of 12 4.3. Authorizations For High Lifetime Exposure 4.3.1. USE Attachment 2, High Lifetime Dose Control Level Extension Form, or a computerized equivalent, to authorize exposures above 1000 mrem for employees with High Lifetime Exposure.

4.3.2. A supervisor from the department requesting approval shall submit a request to the Radiation Protection Department indicating:

The name and identification number of the individual for whom a dose extension is being requested.

A detailed explanation of why the dose extension is necessary.

The requested annual TEDE for the individual.

4.3.3. The Radiation Protection Department shall complete Section II of Attachment 2 or a computerized equivalent.

4.3.4. Written approval is required as follows for any employee who is categorized as having High Lifetime Exposure:

Individuals supervisor and RPM for end of year dose equivalent exceeding 1000 mrem TEDE, not to exceed 2000 mrem TEDE.

Individuals supervisor, RPM, Plant Manager, and Site Vice President for end of year dose equivalent exceeding 2000 mrem TEDE.

4.4. Planned Special Exposures (PSEs)

NOTE: PSEs are not the same as emergency doses. PSEs only apply to adult workers.

4.4.1. PSEs are to be authorized only in exceptional situations when alternatives that might avoid the dose estimated to result from the PSE are not available or are deemed impractical.

4.4.2. A manager from the department requesting a PSE shall submit a request to the Radiation Protection Department, indicating:

The name and identification number of each individual for whom a PSE is being requested, and The nature of the task for which a PSE is being requested, and A detailed explanation of why the PSE is necessary.

4.4.3. Prior written approval is required before the PSE occurs.

RP-AA-203 Revision 3 Page 6 of 12 4.4.4. Prior to participating in a PSE, individuals involved shall be:

1. Informed of the purpose of the PSE.

2. Informed of the estimated dose and associated potential risk or conditions involved in performing the PSE.

3. Instructed in dose reduction measures and techniques for the PSE.

4.4.5. PSE approval is granted when the PSE document is signed (by hand) and dated by:

The individual(s) for whom a PSE is being requested, and The work group manager or other level of supervisory authority for the individual as chosen by the RPM or designee, and The RPM or designee, and The Plant Manager, and The Site Vice President.

NOTE: If there are any periods of exposure during the life of the monitoring individual that have not been determined or documented (i.e., Absent/ No record), then participation in a PSE is not permitted.

4.4.6. All of the individual's previous PSE dose equivalents and previous doses in excess of routine occupational limits must be determined from records for each individual who will participate in the PSE. Doses received in excess of the routine occupational dose limits in effect at the time of exposures during accidents and emergencies must also be determined and subtracted from the limits for PSEs.

4.4.7. DOCUMENT each individual's current year and previous years:

PSE dose equivalents, and Dose equivalents in excess of the exposure limits in effect at the time of the exposures (rows (1) and (2) of Table 1, NRC Exposure Limits, and the former 10 CFR 20.101), and Dose equivalents in excess of any non-NRC exposure limits.

4.4.8. The maximum authorized dose equivalent an individual may receive for a PSE shall be limited to an amount that does not cause the individual to receive a dose equivalent in excess of the limits listed in rows (3) and (4) in Table 1.

4.4.9. DOCUMENT the Planned Special Exposure on Attachment 3, Planned Special Exposure PSE Approval Form, and MAINTAIN all records in accordance with 10CFR20.1205.

4.4.10. SUBMIT a written report to the Administrator of appropriate regional office within 30 days following the PSE in accordance with 10CFR20.1206.

RP-AA-203 Revision 3 Page 7 of 12 4.4.11. SUBMIT a written report of the PSE assigned dose to the individuals involved within 30 days of the PSE.

4.4.12. The dose equivalent received from a PSE is always tracked separately from routine occupational exposure.

4.4.13. Once an exposure is authorized as a PSE, it cannot later be treated as a routine occupational exposure. It must be recorded as a PSE, and all the unique limitations, reporting, and record keeping requirements for PSEs shall apply.

4.5. Emergency Exposure Limits (CM-1) 4.5.1. Emergency exposure in excess of 25 rem TEDE is to be limited to once in a lifetime.

4.5.2. Emergency personnel are to be informed before the fact of possible health effects at the anticipated exposure levels.

4.5.3. For the control of personnel exposures under emergency conditions, LIMIT an individual's dose equivalent per activity as follows:

TABLE 2 - EMERGENCY EXPOSURE LIMITS (REM)

TEDE LDE SDE TODE ACTIVITY 10 30 100 100 Protecting Valuable Property 25 75 250 250 Lifesaving or Protection of Large Populations

> 25 > 75 >250 > 250 Lifesaving or Protection of Large Populations to Workers Fully Aware of the Risks Involved 4.5.4. Emergency exposures shall be voluntary on the part of the involved individual.

4.5.5. CONSULT the Emergency Plan Implementing Procedures regarding approval to exceed NRC exposure limits.

5. DOCUMENTATION 5.1. RETAIN completed exposure authorizations, including Attachments 1, 2, and 3, in accordance with the station records management program. This records program will include appropriate controls for storage and preservation.

RP-AA-203 Revision 3 Page 8 of 12

6. REFERENCES 6.1. Commitments 6.1.1. CM-1 LaSalle Station AIR 1-81-330 regarding provisions for exposures to individuals during an emergency (Section 4.5.).

6.2. User References 6.2.1. 10 CFR 19.13, "Notifications and Reports to Individuals."

6.2.2. 10 CFR 20, "Standards for Protection against Radiation."

6.2.3. USNRC Regulatory Guide 8.7, "Instructions for Recording and Reporting Occupational Radiation Exposure Data, Revision 1, July 1992.

6.2.4. USNRC Regulatory Guide 8.35, "Planned Special Exposures," July 1992.

6.2.5. EPA-400-R-92-001, "Manual of Protective Action Guides and Protective Actions for Nuclear Incidents."

7. ATTACHMENTS 7.1. Attachment 1, Dose Control Level Extension Form.

7.2. Attachment 2, High Lifetime Dose Control Level Extension Form.

7.3. Attachment 3, Planned Special Exposure (PSE) Approval Form.

RP-AA-203 Revision 3 Page 9 of 12 ATTACHMENT 1 Dose Control Level Extension Form Page 1 of 1 Section I: Reason For Extension NAME: ____________________________ SSN: ___________________________

INDIVIDUAL SIGNATURE: _________________________

1. Are other qualified individuals with a lower current year routine TEDE available to perform this work?

Yes___ No___ N/A ___

Remarks___________________________________________________________________

2. State why an extension above 2000 mrem routine TEDE for the year is necessary for this individual.

3. It is requested that the individual named above be permitted to receive a TEDE for the current year of

_____________mrem.

Requestor ____________________________ Date ___/___/___

Section II: Dose Summary CURRENT YEAR ROUTINE RECORD TEDE (mrem): _____________

CURRENT YEAR ROUTINE ESTIMATED TEDE (mrem)*: _____________

CURRENT YEAR ROUTINE TOTAL TEDE (mrem): ______________

LIFETIME DOSE (mrem): ______________

Verified By: _________________ Date: ____________

Section III: Approvals**

The individual named above is approved to exceed 2000 mrem but must remain below 5000 mrem routine TEDE for the current year.

Specific Approval Level in mrem TEDE: ______________

1. Work Group Supervisor _________________________ Date ___/___/___

2. RP Manager __________________________________ Date ___/___/___

3. Station Manager _______________________________ Date ___/___/___

4. Site Vice President _____________________________ Date ___/___/___

• An individual being considered for approval greater than 4000 mrem TEDE for the year should not have significant estimated dose equivalent from a non-ROG facility. Additionally, an individual shall not be approved for greater than 2000 mrem TEDE if that person has any absent/ no record dose equivalent for the year.

• • To raise the ADCL up to and including 3000 mrem, signatures 1 & 2 are required. To raise the ADCL to between 3001 and 4000 mrem, signatures 1,2, & 3 are required. To raise the ADCL above 4000 mrem, signatures 1,2,3, & 4 are required.

RP-AA-203 Revision 3 Page 10 of 12 ATTACHMENT 2 High Lifetime Dose Control Level Extension Form Page 1 of 1 Section I NAME: ____________________________ AGE: _______ SSN: _______________________

INDIVIDUAL SIGNATURE: _________________________

1. State why an extension in excess of the administrative dose control level listed below is necessary for this individual.

2. It is requested that the individual named above be permitted to receive a TEDE for the current year of

_____________mrem.

Requestor ____________________________ Date ___/___/___

Section II: Dose Summary CURRENT YEAR ROUTINE RECORD TEDE (mrem): _____________

CURRENT YEAR ROUTINE ESTIMATED TEDE (mrem):

CURRENT YEAR ROUTINE TOTAL TEDE (mrem): _____________

LIFETIME DOSE (mrem): ______________

Verified By: _________________ Date: ____________

Section III: Approvals The individual named above is approved to exceed 1000 mrem TEDE.

Specific Approval Level in mrem TEDE: ______________

1. Work Group Supervisor _________________________ Date ___/___/___

2. RP Manager __________________________________ Date ___/___/___

3. Station Manager _______________________________ Date ___/___/___

4. Site Vice President _____________________________ Date ___/___/___

• To raise the ADCL to 2000 mrem, signatures 1 & 2 are required.

• • To raise the ADCL above 2000 mrem, signatures 1,2,3 & 4 are required. If approval is being made to exceed 2000 mrem TEDE, this form should be used in conjunction with Attachment 1. Exceptional circumstances should exist to allow individuals with high lifetime dose to exceed 2000 mrem TEDE in a year.

• • • High lifetime exposure controls do not apply to Non-Exelon employees in Midwest due to the limitations of the Exposure Tracking System.

RP-AA-203 Revision 3 Page 11 of 12 ATTACHMENT 3 Planned Special Exposure (PSE) Approval Form Page 1 of 2 Employee Name: ___________________ SSN: _________________

Requested By: _______________________ Date: _________________

Exceptional Circumstances for Scope of Activity:

Actions Necessary:

Justification Why Actions Are Necessary:

Estimated Individual Exposure: _________ Actual Individual Exposure: ___________

Estimated Collective Exposure: _________ Actual Collective Exposure: ___________

Techniques used to maintain exposure ALARA:

Annual Life Time Previous PSE Dose Dose in Proposed Proposed Dose Dose excess of PSE Dose Annual (mrem) (mrem) Life Time Annual limits (mrem) Dose (mrem) (mrem) (mrem) (mrem)

TEDE LDE SDE-ME SDE-WB TODE

RP-AA-203 Revision 3 Page 12 of 12 ATTACHMENT 3 Planned Special Exposure (PSE) Approval Form Page 2 of 2 STATEMENT OF UNDERSTANDING:

I have been informed of the purpose of the planned operation, the estimated doses, and the potential risks or other conditions that may be involved in performing this task. I have been given the opportunity to ask questions and understand the operation and the Planned Special Exposure estimate.

Employee Signature: _________________________________ Date: _______________

Employee Name: _________________________________ SSN: _______________

APPROVALS:

Work Group Manager: _____________________________ Date: _______________

Radiation Protection Manager: _______________________ Date: _______________

Plant Manger: ____________________________________ Date: _______________

Site Vice President: _______________________________ Date: _______________

Secondary Containment Isolation Instrumentation 3.3.6.2 3.3 INSTRUMENTATION 3.3.6.2 Secondary Containment Isolation Instrumentation LCO 3.3.6.2 The secondary containment isolation instrumentation for each Function in Table 3.3.6.2-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.6.2-1.

ACTIONS

---

NOTE-------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Place channel in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for inoperable. trip. Functions 1 and 2 AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for Functions other than Functions 1 and 2 B. One or more Functions B.1 Restore isolation 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with isolation capability.

capability not maintained.

(continued)

Dresden 2 and 3 3.3.6.2-1 Amendment No. 185/180

Secondary Containment Isolation Instrumentation 3.3.6.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1.1 Isolate the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> associated Completion associated Time not met. penetration flow path.

OR C.1.2 Declare associated 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> secondary containment isolation valves inoperable.

AND C.2.1 Place the associated 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> standby gas treatment (SGT) subsystem in operation.

OR C.2.2 Declare associated 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> SGT subsystem inoperable.

Dresden 2 and 3 3.3.6.2-2 Amendment No. 185/180

Secondary Containment Isolation Instrumentation 3.3.6.2 SURVEILLANCE REQUIREMENTS

---

NOTES -----------------------------------

1. Refer to Table 3.3.6.2-1 to determine which SRs apply for each Secondary Containment Isolation Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains isolation capability.

SURVEILLANCE FREQUENCY SR 3.3.6.2.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program SR 3.3.6.2.2 Perform CHANNEL FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program SR 3.3.6.2.3 Calibrate the trip unit. In accordance with the Surveillance Frequency Control Program SR 3.3.6.2.4 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program (continued)

Dresden 2 and 3 3.3.6.2-3 Amendment No. 237/230

Secondary Containment Isolation Instrumentation 3.3.6.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.6.2.5 Perform CHANNEL CALIBRATION. In accordance with the Surveillance Frequency Control Program SR 3.3.6.2.6 Perform LOGIC SYSTEM FUNCTIONAL TEST. In accordance with the Surveillance Frequency Control Program Dresden 2 and 3 3.3.6.2-4 Amendment No. 237/230

Secondary Containment Isolation Instrumentation 3.3.6.2 Table 3.3.6.2-1 (page 1 of 1)

Secondary Containment Isolation Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS TRIP SYSTEM REQUIREMENTS VALUE

1. Reactor Vessel Water 1,2,3, 2 SR 3.3.6.2.1 2.65 inches LevelLow (a) SR 3.3.6.2.2 SR 3.3.6.2.3 SR 3.3.6.2.5 SR 3.3.6.2.6

2. DrywellPressureHigh 1,2,3 2 SR 3.3.6.2.2 1.94 psig SR 3.3.6.2.4 SR 3.3.6.2.6

3. Reactor Building Exhaust 1,2,3, 2 SR 3.3.6.2.1 14.9 mR/hr RadiationHigh (a),(b) SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6

4. Refueling Floor 1,2,3, 2 SR 3.3.6.2.1 100 mR/hr RadiationHigh (a),(b) SR 3.3.6.2.2 SR 3.3.6.2.4 SR 3.3.6.2.6 (a) During operations with a potential for draining the reactor vessel.

(b) During movement of recently irradiated fuel assemblies in secondary containment.

Dresden 2 and 3 3.3.6.2-5 Amendment No. 237/230

Dresden Annex Exelon Nuclear Figure 4-1: Dresden Station PAR Determination Flowchart No Classification is a General Emergency? No PARs Required Yes No Evaluate Dose Assessment results to Is this the Initial PAR?

determine if a PAR upgrade is required Yes Release via Controlled direct containment vent with a Yes duration < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />?

No No Dose Projection < 1 Rem TEDE AND < 5 Rem CDE (thyroid)

Yes Yes Is there a Hostile Action Event in progress?

No Has the utility been informed by the State that there Yes Shelter 2-Mile Radius & 5-Miles Downwind.

are impediments to evacuation?

No Evaluate Dose Assessment results to Evacuate 2-Mile Radius & 5-Miles Downwind determine if a PAR upgrade is required.

2 Mile Radius, 5 Miles Downwind WD (from) Subareas 002º to 046º 1, 3. 4, 7 047º to 182º 1, 3. 4 183º to 292º 1, 3, 4, 12 293º to 299º 1, 3, 4 300º to 338º 1, 3, 4, 9 339º to 001º 1, 3, 4, 7, 9 June 2013 DR 4-4 EP-AA-1004 (Revision 33)

DEOP 200-1 PRIMARY CONTAINMENT CONTROL ENTRY CONDITIONS Drywell temperature Torus bulk Torus water level Drywell or torus Drywell pressure above 160°F temperature below -4.5 in.

hydrogen above 2.0 psig (Use TR 2(3)-1340-1 OR above 95°F above 3.5%

Points 5 and 6) above -1.5 in.

PRIMARY CONTAINMENT DRYWELL TORUS TEMPERATURE TORUS WATER LEVEL HYDROGEN PRESSURE TEMPERATURE Hold drywell and torus pressures below 2.0 psig using Hold drywell temperature below 160°F using drywell cooling. CAUTION: Exceeding LPCI NPSH/Vortex Hold torus water level between -4.5 in. and -1.5 in. narrow range. Monitor hydrogen and oxygen concentrations in drywell and torus SBGT and drywell purge (DOP 1600-1). Limits (Figs V-X) may cause (DOP 2400-1).

Drywell temperature affects RPV water level indication. system damage. Sample the torus before discharging water.

Check Detail A.

IF THEN IF THEN IF THEN IF THEN Cannot hold pressure below Go to t. Hydrogen or oxygen monitor is Sample the drywell and torus for Cannot hold drywell Go to i. Low Level: Cannot hold level Go to s.

2.0 psig Hold torus bulk temperature below 95°F using torus cooling. unavailable hydrogen and oxygen.

temperature below 160°F above -4.5 in.

High Level: Cannot hold level Go to a. ENTER HYDROGEN CONTROL:

IF THEN below -1.5 in.

Hydrogen (at or above 1%) or oxygen (at or above 5%) is detected Enter DEOP 200-2 while t i Cannot hold torus bulk Go to o. in the drywell or torus continuing in other sections £ temperature below 95°F of this procedure.

BEFORE Start all available drywell cooling Drywell pressure reaches (Torus spray) Hydrogen or oxygen concentration in ENTER HYDROGEN CONTROL:

OK to defeat drywell cooling 9 psig isolations (DEOP 500-2). High Level (Above -1.5 in.) a the drywell or torus is unknown Enter DEOP 200-2 while o continuing in other sections £ of this procedure.

Start all available torus cooling Hold torus water level below 18.5 ft (ring header).

BEFORE Do not use pumps needed IF THEN for core cooling. IF THEN Drywell temperature reaches (Drywell spray)

Torus sprays running Before torus pressure drops to 0 psig, 281°F Cannot hold level below 18.5 ft 1. Scram.

stop torus sprays. (ring header)

2. ENTER RPV CONTROL: £ BEFORE Drywell sprays running Before drywell pressure drops to 0 psig, Enter DEOP 100 stop drywell sprays. (Scram, Enter DEOP 100) while continuing here Torus bulk temperature reaches IF THEN 110°F 3. Stop injection from outside the primary containment not needed Drywell sprays running Before drywell pressure drops to 0 psig, for core cooling or to shut down stop drywell sprays. the reactor.

Torus water level At or above 27.5 ft 4. IF .........you cannot restore torus

? 1. Scram. water level below 18.5 ft and hold it there,

2. ENTER RPV CONTROL: THEN ..BLOW DOWN:

Enter DEOP 100 Enter DEOP 400-2 Below 27.5 ft. Below No while continuing here. £ while continuing here.

£ Fig K, Drywell Spray Initiation Limit CAUTION: Exceeding LPCI NPSH/Vortex Limits  ?

(Figs V-X) may cause system damage.

Yes Start torus sprays. Hold torus bulk temperature below Fig M, Heat Capacity Limit.

Do not use pumps needed for core cooling.

Low Level (Below -4.5 in.) s CAUTION: Exceeding LPCI NPSH/Vortex Limits OK to use external spray sources.

(Figs V-X) may cause system damage. IF THEN Reducing primary containment pressure affects margin to Hold torus water level above 12 ft. (HPCI exhaust).

NPSH limits. Check Figs W-Y, NPSH Limits. 1. Trip all recirc pumps. 1. IF......... you are not in RPV Flooding Cannot hold torus bulk temperature

2. Trip all drywell cooling fans. below Fig M, Heat Capacity Limit (DEOP 400-1) or Steam Cooling (DEOP 400-3), IF THEN

3. Start drywell sprays. THEN .. lower RPV pressure to stay Do not use pumps needed for core cooling. below Fig M, Heat Capacity Cannot hold level above 12 ft. Trip HPCI.

Limit. (HPCI exhaust) Even if core cooling will be lost.

OK to use external spray sources if you can restore and WAIT until hold torus bottom pressure and primary containment water OK to exceed 100°F/hr drywell pressure level inside Fig D, Primary Containment Pressure Limit. cooldown rate.

is above Cannot hold level above 11 ft. 1. Scram.

Reducing primary containment pressure affects margin 2. IF......... you still cannot stay below (downcomers) 9 psig to NPSH limits. Check Figs W-Y, NPSH Limits. Fig M, Heat Capacity Limit, 2. ENTER RPV CONTROL:

THEN .. BLOW DOWN: Enter DEOP 100 Enter DEOP 400-2 while continuing here £ while continuing here. £

3. BLOW DOWN:

Keep trying to lower drywell temperature below 160°F. Enter DEOP 400-2 Below while continuing here

£ No Drywell temperature affects RPV water level indication.

Fig K, Drywell Spray Initiation Limit Check Detail A.

?

Yes IF Cannot restore drywell 1. Scram.

THEN A RPV Water Level Instruments temperature below 281°F CAUTION: RPV water level instruments may be unreliable due CAUTION: Exceeding LPCI NPSH/Vortex Limits 2. ENTER RPV CONTROL:

and hold it there to boiling in the instrument runs when drywell or (Figs V-X) may cause system damage. Enter DEOP 100

£ while continuing here reactor building temperatures near the instrument

1. Trip all recirc pumps. runs are above Fig B, RPV Saturation Temperature.

3. BLOW DOWN:

2. Trip all drywell cooling fans. Enter DEOP 400-2 An RPV water level instrument may not be used if the instrument while continuing here

£ reads at or below its Minimum Usable Level (Detail C).

3. Start drywell sprays.

Do not use pumps needed for core cooling.

OK to use external spray sources if you can restore and B RPV Saturation Temperature hold torus bottom pressure and primary containment water level inside Fig D, Primary Containment Pressure Limit. 600 Reducing primary containment pressure affects margin Drywell Temperature (°F) to NPSH limits. Check Figs W-Y, NPSH Limits.

500 D Primary Containment Pressure Limit K Drywell Spray Initiation Limit L Pressure Suppression Pressure M Heat Capacity Limit TR 2(3)-1340-1, Point 9 or 10 70 600 40 230 400 Keep trying to lower drywell and torus pressures below OR 9 psig.

220 60 35 Reactor Building Temperature (°F)

IF THEN 500 210 300 Drywell Temperature (°F)

Torus Bottom Pressure (psig) Torus Bottom Pressure (psig) 30 Torus Bulk Temperature (°F)

Cannot stay inside Fig L, Go to y. 50 200 Pressure Suppression 25 66 psig Pressure 400 190 200 40 0 100 200 300 400 500 600 700 800 900 1000 1100 20 180 RPV Pressure (psig)

TR 2(3)-1340-1 Points 5 and 6 30 300 170 15 y

20 160 Use only when torus level is at or below 17 ft. C Minimum Usable Indicating Levels 10 See TSGs for torus levels above 17 ft.

BLOW DOWN: 150 200 Drywell temperature (°F)

Enter DEOP 400-2 10 TR 2(3)-1340-1, Point 9 or 10 while continuing here. £ 5 140 32 101 201 301 401 501 to to to to to to 0 100 0 130 0 10 20 30 40 50 60 70 80 90 100 100 200 300 400 500 558 0 2 4 6 8 10 12 14 16 18 20 10 11 12 13 14 15 16 17 18 19 20 0 100 200 300 400 500 600 700 800 900 1000 1100 Primary Containment Water Level (ft) 93 10.9 18.6 BEFORE Drywell Pressure (psig) Wide Range Torus Water Level (ft) RPV Pressure (psig) Fuel Zone

(+60" to -340") -297 -297 -298 -299 -300 -301 Torus bottom pressure reaches (Vent)

Fig D, Primary Containment LPCI / Core Spray NPSH Limit LPCI / Core Spray NPSH Limit Medium/Narrow Range Pressure Limit V ECCS Vortex Limit W ECCS Flow Up To 10,750 gpm X ECCS Flow Up To 25,500 gpm Y HPCI NPSH Limit All Rx Bldg Temps

(+60" to -60")

-38 -39 -41 -43 -43 -43 11 250 250 250 Medium/Narrow Range Rx Bldg Temps

-60 -60 -60 -60 -60 -60 181°F or less Vent to stay below Fig D, Primary Containment Pressure (+60" to -60")

Limit (DEOP 500-4).

Wide Range Torus Water Level (ft)

Torus Bulk Temperature (°F) Torus Bulk Temperature (°F) Torus Bulk Temperature (°F) 10.5 200 200 200 Wide Range OK to exceed release rate limits. -68 -51 -21 17 68 107

(+330" to -70")

15 psig 15 psig Torus Bottom Pressure (psig) Torus Bottom Pressure (psig) 10 150 150 150 10 psig 10 psig Torus Bottom Pressure (psig) 15 psig 5 psig 5 psig 9.5 100 100 100 10 psig 5 psig 3.5 psig CATEGORY 1 3.5 psig 3.5 psig Dresden Nuclear Power Station Units 2(3) 9 50 50 50 EMERGENCY OPERATING PROCEDURE 0 5000 10,000 15,000 20,000 25,000 30,000 35,000 0 1000 2000 3000 4000 5000 6000 0 1000 2000 3000 4000 5000 6000 0 1000 2000 3000 4000 5000 6000 7000 LPCI / Core Spray Pump Flow (gpm) LPCI / Core Spray Pump Flow (gpm) HPCI Flow (gpm) Title Total ECCS Flow (gpm)

PRIMARY CONTAINMENT CONTROL Number Rev DEOP 200-1 10

DEOP 300-1 DEOP 300-2 A RPV Water Level Instruments SECONDARY CONTAINMENT CONTROL RADIOACTIVITY RELEASE CONTROL CAUTION: RPV water level instruments may be unreliable due to boiling in the instrument runs when drywell or reactor building temperatures near the instrument runs are above Fig B, RPV Saturation Temperature.

An RPV water level instrument may not be used if the instrument reads at or below its Minimum Usable Level (Detail C).

ENTRY CONDITIONS ENTRY CONDITIONS B RPV Saturation Temperature Any of the following in the reactor building:

600 Off-site release rate Area temperature Differential Area radiation Floor drain sump Corner room above GSEP Alert level (Dresden EALs)

Vent radiation above max normal Pressure above max normal water level water level Drywell Temperature (°F) above 4 mr/hr (Detail S) at or above 0 in. (Detail T) above hi-hi alarm above 0 in.

500 TR 2(3)-1340-1, Point 9 or 10 400 OR IF THEN

• Operate Turbine Building Ventilation.

Reactor building exhaust radiation above 4 mr/hr Verify:

• Isolate all primary system discharges outside primary and secondary Reactor Building Temperature (°F) 300

• Reactor Building Ventilation isolation containments except systems needed for other DEOP actions.

• SBGT start IF THEN 200 Reactor Building Ventilation isolates Start Reactor Building Vent:

Primary system discharging outside Before off-site release rate reaches the 0 100 200 300 400 500 600 700 800 900 1000 1100 AND

• OK to defeat high drywell pressure and low RPV water level interlocks primary and secondary containments General Emergency level RPV Pressure (psig) Reactor building exhaust radiation below 4 mr/hr (DEOP 500-2). AND (Dresden EALs):

Discharge cannot be isolated

1. Scram.

C Minimum Usable Indicating Levels 2. ENTER RPV CONTROL:

Enter DEOP 100 Drywell temperature (°F) CAUTION: High temperature or low water level in the spent fuel pool can while continuing here.

TR 2(3)-1340-1, Point 9 or 10 affect radiological conditions and water levels in the reactor 32 101 201 301 401 501 building and adjacent areas. Reactor building access may also be 3. BLOW DOWN:

to to to to to to restricted by steam. Enter DEOP 400-2 100 200 300 400 500 558 while continuing here.

Fuel Zone

(+60" to -340") -297 -297 -298 -299 -300 -301 Medium/Narrow Range All Rx Bldg Temps -38 -39 -41 -43 -43 -43

(+60" to -60")

Medium/Narrow Range Rx Bldg Temps

-60 -60 -60 -60 -60 -60 TEMPERATURE / RADIATION WATER LEVELS 181°F or less

(+60" to -60")

Wide Range -68 -51 -21 17 68 107

(+330" to -70") Operate sump pumps to hold floor drain sump level below Operate area coolers and Reactor Building Ventilation.

the hi-hi alarm setpoint and remove water from reactor

• Reactor building temperature affects RPV water level building areas.

indication. Check Detail A.

IF THEN IF THEN Floor drain sump level cannot Go to j Any area temperature or Go to j be restored below the hi-hi radiation above max normal alarm setpoint (Details S, T) OR S Reactor Building Area Temperatures Water cannot be removed from an area Max Normal Max Safe Area Temperature (°F) Temperature (°F)

HPCI Room 150 210 Shutdown Cooling Pump Room 150 180 j Isolate all discharges into affected areas Shutdown Cooling Ht X Room 150 180 except systems needed for:

• Fire fighting Clean Up Demin Room 150 210

• Other DEOP actions Clean Up Pump & Ht X Area 150 210 Isolation Condenser Area 150 180 IF THEN Primary system discharging into Go to k reactor building AND Discharge cannot be isolated T Reactor Building Area Radiation Levels k

Max Normal Max Safe Area Radiation (mr/hr) Radiation (mr/hr) BEFORE WAIT until Any area temperature, radiation, or HPCI Cubicle, Unit 2(3) 150 (100) 2500 2 or more areas (Scram, Enter DEOP 100) water level reaches max safe value above max safe values (Details S, T, U) of same parameter East LPCI Pump Area 12 2500 * (Details S, T, U)

West LPCI Pump Area 9 2500

• East CRD Module Area 30 2500

• 1. Scram.

2. ENTER RPV CONTROL:

West CRD Module Area 50 2500

• Enter DEOP 100 Shut down the reactor. while continuing here.

Vessel Instrument Rack Area 30 2500

• Clean Up System Area 30 2500

• Isolation Condenser Area, Unit 2(3) 10 2500 (2500 * ) WAIT until 2 or more areas above max safe values

• Measured by local survey. of same parameter (Details S, T, U)

Reactor Building U Area Water Levels Max Safe Area Water Level (in.) BLOW DOWN:

East Corner Room Floor 8 Enter DEOP 400-2 while continuing here.

CATEGORY 1 CATEGORY 1 Dresden Nuclear Power Station Units 2(3) Dresden Nuclear Power Station Units 2(3)

West Corner Room Floor 8 EMERGENCY OPERATING PROCEDURE EMERGENCY OPERATING PROCEDURE Title Title Secondary Containment Control Radioactivity Release Control Number Rev Number Rev DEOP 300-1 09 DEOP 300-2 02

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