ML14010A308

Final Safety Analysis Report, Amendment 62, Chapter 12 - Radiation Protection, Figure 12.3-19 Through Page 12.5-21

ML14010A308
Person / Time
Site:	Columbia
Issue date:	12/30/2013
From:	Energy Northwest
To:	Office of Nuclear Reactor Regulation
Shared Package
ML14010A476	List: GO2-13-174, Final Safety Analysis Report, Amendment 62, Appendix J - Shielding Evaluation Report ML14010A290 ML14010A291 ML14010A292 ML14010A293 ML14010A294 ML14010A295 ML14010A296 ML14010A297 ML14010A298 ML14010A299 ML14010A300 ML14010A301 ML14010A302 ML14010A303 ML14010A304 ML14010A305 ML14010A306 ML14010A307 ML14010A308 ML14010A309 ML14010A310 ML14010A311 ML14010A312 ML14010A313 ML14010A314 ML14010A315 ML14010A316 ML14010A317
References
GO2-13-174
Download: ML14010A308 (51)

Text

Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Figure Not Available For Public Viewing Amendment 54 April 2000 Arrangement of Filtration and DemineralizationEquipment (Typical) 900547.52 12.3-19 Figure Form No. 960690Draw. No.Rev.Valve Gallery El. 487' - 0" El. 507' - 0" El. 467' - 0" Corridor Filter Demineralizer Instrument Rack 2'-0" MinValve Reach Rod Corridor Filter Holding PumpWall Penetrations

Not in Line of Sight

of Radiation Source Removable Shield PlugsSteel Plate (Typ.)Wall Sleeve Penetration Detail Grout Columbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000Schematic Arrangement of the Cooler Condenser Loop Seal 900547.53 12.3-20 Figure Form No. 960690Draw. No.Rev.Notes:1. Valves and instrumentation are not shown to prevent clutter.

2. Overflow volume in the enlarged discharge section is sufficient to restore loop seal following a pressure surge.Off-gas Stream See Note 2To Sump 16'Seal(Typ.)Cooler Condenser Glycol Loop(Typ.)Moisture Separator Columbia Generating StationFinal Safety Analysis Report Amendment 54 April 2000 Figure Form No. 960690Draw. No.Rev.Decontamination Concentrator Steam Supply Arrangement 900547.54 12.3-21 DistillateTank Evaporator Cond ReturnChem Waste Concentrate

Discharge Auxiliary Steam Concentrator Heating Element Columbia Generating StationFinal Safety Analysis Report Figure Not Available For Public Viewing Amendment 54 April 2000 6' - 4"Layout of the Standby Gas Treatment SystemFilter Units 900547.56 12.3-23 Figure Form No. 960690Draw. No.Rev.4.1 H.3 J K 5 5.2 6 6.8 7.7 8.3 5' - 0" Stand-by GasTreatment Unit 1A Dop and Freon Injection Port Moisture Separator Electric Heating Coils PrefilterHigh Efficiency Filter (HEPA)Carbon Test Canisters(Total 12)

Stand-by GasTreatment Unit 1B Activated Carbon Iodine AdsorberCarbon Test Canisters(Total 12)High Efficiency Filter (HEPA)Access Door 20x50 (Typical)Butterfly Damper (Typical)

Dop and Freon Detection PortExhaust Fan with AutomaticInlet Vanes (Typical)

Reactor Building El. 572' - 0" Columbia Generating StationFinal Safety Analysis Report

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 LDCN-05-024 12.4-1 12.4 DOSE ASSESSMENT

12.4.1 DESIGN CRITERIA

The criteria for the dose to plant personnel during normal opera tion and anticipat ed operational occurrences including refueling, are based on the requirements discussed in 10 CFR Part 20.

The design radiation levels during normal operation and refueling are shown in Figures 12.3-5 through 12.3-18. In areas such as the control room and offices, the maximum dose rate does not exceed 1.0 mrem/hr (Zone I radiation level). For pers onnel who work in controlled radiation areas, radiation Zone II through IV in Figures 12.3-5 through 12.3-18, administrative controls ensure that doses do not exceed the requirements of 10 CFR Part 20.

12.4.2 PERSONNEL DOSE ASSESSMENT BASED ON BWR OPERATING DATA

The italicized information is historical and was provided to support the application for an operating license.

12.4.2.1 General

In general, data (Reference 12.4-1) from operating boiling water reactors (BWRs) have shown that the man-rem exposures to plant personnel are primaril y due to the corrosion product isotopes. Of the corrosion product isotopes, 60Co is believed to be the single most important radionuclide.

A review of the data from operating re actors was performe d in References 12.4-6 and 12.4-7. Based on this it was concluded that the shield ing design, which assumes the GE BWR design base source terms, was adequate to account for the additional radi oactivity that will deposit in the lines due to crud.

Chemical cleaning connections were also installed on a number of systems. A chemical cleanup can be performed to reduc e the deposits of crud and minimi ze the increase in radiation levels if needed. Section 12.3.1.3.2 addresses the design features that were incorporated to reduce the buildup of crud.

The variables that have been found to affect plant personnel exposure in clude the following:

a. The BWR plants show an increase in total personnel exposure during the first few years of operation,

b. The need to minimize plant downtime requires that inspection and repair tasks must be started immediately after plant shutdown when the dose rates from short-lived radionuclides can be significant, C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 12.4-2 c. Plant design and equipmen t layout has a significant e ffect on personnel dose.

Section 12.3.1 discusses the design features used to mi nimize plant personnel exposure, d. Training and experienc e of plant workers,

e. The extent of maintenance operations required for a specific year, and

f. The extent that a utility uses non-regular or contractor personnel.

12.4.2.2 Personnel Dose fr om Operating BWR Data

References 12.4-1 through 12.4-5 provide a tabulation of pers onnel exposures for operating BWRs. Table 12.4-9 tabulates the average personnel exposure for operating BWRs for the period 1969 through 19

80. References 12.4-4 and 12.4-5 provide more recent information.

The assessments of personnel ex posures summarized in Section 12.4.2.3 include this more recent information.

12.4.2.3 Occupancy Factors, Dose Ra tes, and Estimated Personnel Exposures

A summary of the total estimated man-rem doses broken down by major function is given in Table 12.4-1. More detailed break downs are presented in Tables 12.4-2 through 12.4-8 for each of the seven major functions given in Table 12.4-1. These tables are based on the more recent information obtained from indus try operating experience. The data from Table 12.4-9 is given for comparison purposes only.

The results of the to tal estimated man-rem doses will be discussed with reference to six occupational groups as follows:

a. Group 1 - This group includes mainte nance personnel such as mechanical, electrical, instrument cr aftsmen, and Foreman.

There are approximately 128 people is this group.

Tables 12.4-4 and 12.4-8 provide the functional breakdown of exposures for this occupational group. As can be seen from the tables, 433 total man-rem may be expected.

Routine and special mainte nance operations which include control rod drive repairs, residual heat removal (RHR) repairs, snubber maintenance, etc., account for approximately 60% of the average annual personnel dose. One to two rem per year per person is projected for the station maintenance personnel for a maximum total of 256 man-rem per year. Accordingly, the remaining 175 man-rem per year would be expected to be received by non-station maintenance personnel. As discussed in Section 12.3.1 , the equipment layout C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 12.4-3 and design and shielding design are such that the exposur es are as low as is reasonably achievable (ALARA).

b. Group 2 - This group includes plant operations personnel composed of supervisors, control room staff and plant equipment operators. There are approximately 60 people in this group.

Tables 12.4-2 , 12.4-3 , 12.4-5 , and 12.4-6 show the total estimated man-rem fo r this group. As can be seen, the total is approximately 130 man-rem per year or approximately 2.2 rem per year per man. Personnel in this group will be performing routine and non-routine operation and surveillance, waste proces sing and refueling opera tions. In plant operations, personnel are expected to rece ive approximately one to two rem per year per man for a maximum total of 120 man-rem per year. The remaining 10 man-rem per year may be expected to be received by non-station personnel.

As part of this total, the supervisors and control room sta ff are expected to receive an exposure of less than 500 mrem/yr.

c. Group 3 - This group includes health physics and chemistry personnel. There are approximately 53 people in this group.

If the plant chemistry personnel spend 1% of their time collecting samp les in Zone III sampling stations. They will receive a maximum dose of 723 mrem/yr. Assuming the remainder of their time is spent in Zone I and Zone II areas, the total dose is between 1 and 2 rem per person. The health physics pers onnel conduct radiation surveys and support maintenance activities which require con tinuous and pre-job radi ation surveys.

The exposure to these health physics personnel ranges from 2 to 3 rem/yr. This is based on experience from operating pl ants. Assuming a dose of 3 rem per person per year and consider ing 35 health physics people in the group, the total is 105 man-rem per year. Since this group covers virtually all functions

delineated in Tables 12.4-2 through 12.4-8, this 105 man-rem is considered to be spread out across all the functions.

d. Group 4 - This group includes engineers and technical supervisors. There are approximately 27 people in this group.

Personnel in this group will spend most of their time in Zone I areas where exposures are less than 500 mrem/yr. Table 12.4-7 indicates approximately 153 man-rem per year will be experienced for inservice inspection.

Plant technical personnel will have a supervising roll in this operation with non-stati on personnel performing the inspection operations. This supervisory roll will take the personnel into all zone levels during ISI activities and this roll is expec ted to result in exp osure from 1 to 2 rem/yr. Thus, the projected dose estimate for the 27 people in this group is 54 man-rem per year, the balance bei ng accounted for in the non-station personnel.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 12.4-4 e. Group 5 - This group includes station s upervisors such as health physics and chemistry supervisors, shift supervisors, etc. There are approximately 24 people in this group. Station personnel will supervise Group 1 and Group 2 personnel.

Their dose is approximately the same as personnel in these groups. With a projected dose estimate of 1 rem per year per person with 24 people in the group, the total dose is 24 man-rem per year.

f. Group 6 - This group includes adminis trative and manageme nt personnel.

There are approximately 31 people in this group. Personnel in this group spend their time in Zone I radiation areas. The projected dose estimates will be less than 500 mrem/yr. With 31 people in this group and a 500 mrem per man per year the total dose is 15.5 man-rem per year.

As seen from Table 12.4-1 , the total estimated man-rem expos ure is 715 man-rem. Groups 3, 5, and 6 are considered to be spread over all the functions. These groups constitute only 15%

of the total exposure in any case.

12.4.3 INHALATION EXPOSURES

Airborne radionuclide concentrations in normally occupied areas are, as discussed in

Section 12.2.2 , well below the limits set by 10 CFR Part 20 and thus inhala tion exposures are negligible. In areas where engineering contro ls or operational procedures do not reduce the airborne radionuclide concentrations sufficiently, additional measures such as access control, limiting exposure time (DAC hours), and respirator y protection devices are used to maintain the total effective dose e quivalent (TEDE) ALARA.

12.4.4 SITE BOUNDARY DOSE

Steam handling equipment on the turbine operating floor can contribute to the site boundary

dose in two ways: through a direct component and through an air-scattered "skyshine" component. Since the 16N bearing equipment is known, it can be shielded to reduce the direct component. The "skyshine" component reaches the site boundary as a result of those gamma rays which are directed such th at they bypass any inte rcepting shield walls and are scattered by the air to the site boundary.

The calculated results show that the skyshine dose will have its greatest effect on a dose point approximately 1950 m north of the turbine buildi ng. The skyshine dose at this point will be approximately 4 mrem/yr. This result is based on a plant capacity factor of 80%.

The main contributors to this dose and their contribution (in percent) are the south moisture-separator reheater (MSR) which contributes 60%, the north MSR which contributes 20%, the cross over lines which contribute 10%, and the turbines and feedwater heaters which contribute 10%.

C OLUMBIA G ENERATING S TATION Amendment 58 F INAL S AFETY A NALYSIS R EPORT December 2005 12.4-5 The dose estimate was computed from a model that represents the 16 N gamma leakage by point isotopic sources. This model uses the out put from the COHORT Code (Reference 12.4-3) which gives the airscattered dose as a func tion of distance a nd source ray angle.

The site boundary dose from liquid and gase ous effluents are disc ussed in Sections 11.2.3 and 11.3.3.

12.

4.5 REFERENCES

12.4-1 Atomic Industrial Fo rum, Compilation and Analysis of Data on Occupational Radiation Exposure Experienced at Oper ating Nuclear Power Plants, September 1974.

12.4-2 Ninth Annual Occupati onal Radiation Exposure Report, NRC, NUREG-0322, Washington, D.C., October 1977.

12.4-3 Tenth Annual Occupational Radiation Exposure Report, NRC, NUREG-0463, Washington, D.C., October 1978.

12.4-4 Occupational Radiation Exposure at Light Water Cooled Power Reactors, Annual Report 1977, NRC, NUREG-0482, Washington, D.C., April 1977.

12.4-5 Occupational Radiation Exposure at Commercial Nuclear Power Reactors, Annual Report 1979 and 1980, Volumes 1-2, NRC, NUREG-0713, December 1981.

12.4-6 NRC Seventh Annual Occupati onal Radiation Exposure Report 1974, NUREG-75/108, November 1975.

12.4-7 Atomic Industrial Fo rum, Compilation and Analysis of Data on Occupational Radiation Exposure Experienced at Operating Nuclear Power Plants, September 1974.

C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-7 Table 12.4-1 Summary of Occupati o nal Dose Estimates Man-rem/yr

1. Routine op e ration and s u rveillance 53 2. Nonroutine operation and surveill a nce 15 3. Routine ma i n tenance 288

4. Waste processing 15 5. Refueling 48 6. Inservice inspection 153 7. Special maintenance 145 Total 717 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-8 Table 12.4-2 Occupational Dose E s timates During Routine

Operations and Surveillance Activity Average Dose Rate (mrem/hr)

Exposure Time (hr)

Number of Workers Frequency Dose (man-rem/y r) 1. Walking 0.5 0.5 2 1/shift = 0.54 2a. Checking Railroad acce s s Change rooms Relay room

Motor generator sets Battery room Computer ro om Switch gear r o om Air conditi o ni ng equip. Recirc. motor gen.

RBCCW heat Exchangers Emergency air comp RBCCU pumps RBCCW expansion

Tank 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift = 2.2 2b. Mech. vac. pumps CRD pumps

CRD hydraulic Cont. units Refueling floor CRD filters RUCV demmo resin Tanks RNP pumps SRMP pumps

Air coolers IVST racks 10 10 10 10 10 10 10 10 10 10 10 10 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 2 2 2 2 2 2 2 2 2 2 2 2 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift = 11 2c. CRD storage and repair SGTS HPCI turbine and pump 15 15 15 0.2 0.2 0.2 2 2 2 1/shift 1/shift 1/shift = 3.3 2d. RWCU heat exchangers RHR heat exc h angers Acid purple and turbine 50 50 50 0.1 0.1 0.1 1 1 1 1/shift 1/shift 1/shift = 5.5 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-9 Table 12.4-2 Occupational Dose E s timates During Routine

Operations and Surv e illance (Continued)

Activity Average Dose Rate (mrem/hr)

Exposure Time (hr)

Number of Workers Frequency Dose (man-rem/y r) 2e Checking (continued)

Demin precoat tank Precoat pump Waste sample pump Floor drain s a mple room

Waste surge pump Equip. drain s u mp pump Waste surge pump

Waste precoat pump Waste sludge d i sch. pump Waste filter aid pump Chemical waste pump Floor drain co l l. pump 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.2 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 1 1 1 1 1 1 1 1 1 1 1 1 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift = 0.1 2f. Chemical waste tank Spent resin p u mp Cond. phase d e cant pump

Cond. phase sludge Discharge mix i ng pump 50 50 50 50 50 0.5 0.5 0.5 0.5 0.5 1 1 1 1 1 1/shift 1/shift 1/shift 1/shift 1/shift = 1.3 2g. Floor drain d e min. Waste hopper Floor drain fi lt er 8 8 8 2 2 2 1 1 1 1/shift 1/shift 1/shift = 0.8 2h. Turbine inst. and controls Gen. C O 2 units Station air co m p. Heater feed p u mps Demin. pumps and valves MTG lubricati o n system Hatch area a bo ve demin.

tanks H 2 seal 2.1 eq u i p. Health shell pu l l space BCCW heat expansion

and pumps 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 1 1 1 1 1 1 1 1 1

1 2 2 2 2 2 2 2 2 2

2 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift = 1.1 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-10 Table 12.4-2 Occupational Dose E s timates During Routine

Operations and Surv e illance (Continued)

Activity Average Dose Rate (mrem/hr)

Exposure Time (hr)

Number of Workers Frequency Dose (man-rem/y r) 2i Checking (continued)

TBCCW expansion tank Ventilation equipment Demin. precoat and resin tanks Demin. precoat pumps

Sump pumps Reactor feed pump turbine

Lub. system MTG lub oil cooler Main gen. and exciter

MTG utilizer activators Stop and throttle valves

Circ. water isol. valves 5

5 5

5 5

5 5

5 5

5 5

5 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 0.3 1 1

1 1

1 1

1 1

1 1

1 1 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift = 1.6 2j. Heater drain pumps Heater drain flash tanks Condense water box Reactor feed pumps and turbines 50 50 50 50 0.2 0.2 0.2 0.2 1 1

1 1 1/shift 1/shift 1/shift 1/shift = 11.0 2k. Drain coolers Feed water heaters Reheater seal tank Gland steam condenser

Main turbine

Reheater separators 15 15 15 15 15 15 0.5 0.5 0.5 0.5 0.5 0.5 1 1

1 1

1 1 1/shift 1/shift 1/shift 1/shift 1/shift 1/shift = 14.6 Total 5 3.04 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-11 Table 12.4-3 Occupational Dose Est i mates During Nonroutine

Operations and Surveillance Activity Average Dose Rate (mrem/h r) Exposure Time (hr)

Number of Workers Frequency Dose (man-rem/y r) 1. 1a.

1b.

1c.

1d. Operation of e q uipment:

Traversing in-core probe system

Safety i n jection system Feedwater pumps and turb ine Instrument cal i bration 2 5

1 2 2 1

1 1 2 1

1 1 3/yr 1/month 1/week 1/day 0.02 0.06 0.05 0.73 2. Collection of r a dioactive sam p l es: 2a. 2b.

2c.

2d.

2e.

2f. Liquid system

Gas system

Solid system

Radiochemistry

Radwaste operation

Health physics 10 5 10 1 3

5 0.5 0.5 0.5 1 8

2 1 1

1 2

3 2 1/day 1/month 4/yr 1/day 1/week 1/day 1.83 0.03 0.01 0.73 3.75 7.30 Total 14.50 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-12 Table 12.4-4 Occupational Dose E s timates During Routine

Operations and Surveillance Activity Average Dose Rate (mrem/hr)

Exposure Time (hr)

Number of Workers Frequency Dose (man-rem/y r) 1. 2.

3.

4. 5. 6. 7.

8.

9.

10.

11.

12.

13.

14.

15.

16.

17.

18.

19.

20.

21.

22.

23.

24.

25.

26.

27.

28. Minor repairs reactor building Ventilation and air conditioning Control rod drive repair Reactor water cleanup pump Reactor water cleanup valve and heat exc h anger Residual heat removal system

Safety relief valves

Main steam isol. valves

Recirc. pumps Snubber inspector and repair

Misc. turbine bldg. repairs Reactor feed pumps and turbine

Drain coolers

Steam jet air ejectors

Offgas system

MTG actuator Heater drain flash tanks

Condenser water box Annual turbine inspection Misc. radwaste pump repairs

Misc. radwaste valve repairs Filter and demin.

Centrifuge

Evaporation

Turbine instr. and control Waste solidification

Area monitors Operate laundry facility 1 0.5 15 180 110 200 80 75 200 75 2 10 2 10 2 5 2 5

3 25 10 65 50 85 2 2 20 0.5 20 20 200 35 45 27 30 100 50 100 8 40 40 40 40 40 40 20 120 40 40 30 8 50 10 40 40 40 2 1

6 3 6 8 5 6

3 5

1 2

2 2

2 1

1 1 10 2 2

3 2

3 1

2 2

3 1/week 1/week 1/yr 1/yr 1/yr 1/yr 1/yr 1/yr 1/yr 1/yr 1/day 2/yr 1/yr 2/yr 6/yr 1/yr 1/yr 1/yr 1/yr 4/yr 6/yr 1/yr 4/yr 1/yr 1/week 2/yr 2/yr 1/day 2.1 0.5 18 19 30 43 12 45 30 37.5 5.8 0.8 0.16 1.6 0.96 0.24 0.08 0.1 3.6 8.0 4.0 5.9 3.2 12.8 1.0 0.32 0.32 2.2 Total 288.2 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-13 Table 12.4

-5 Occupational Dose Estim a tes During Waste Processing Activity Average Dose Rate (mrem/hr)

Exposure Time (hr)Number of Worke r s Frequency Dose (man-rem/yr) Radwaste c o ntrol room

.5 8 1 1/shift 4.4 Sampling and filter changing 15 8 1 1/week 6.2 Panel operator insp. and testing 1 2 1 1/day 0.73 Operation of waste and

packaging equipment 2 16 2 1/week 3.3 Total 14.6 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-14 Table 12.4-6 Occupational Dose Estimates During Refueling

Activity Aveg. Dose Rate (mrem/hr)

Exposure Time (hr)

Number of Workers Frequency Dose (man-rem/yr) 1. Opening/closing reactor pressure vessel 60 40 10 1/yr 24 2. Fuel preparation 10 24 2 1/yr 0.48 3. Refueling 10 100 15 1/yr 15 4. Fuel handling 2.5 100 4 1/yr 1.0 5. Fuel sipping 10 120 6 1/yr 7.2 Total 47.7 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-15 Table 12.4-7 Occupational Dose Estimat e s During Inservice I n spection Activity Average Dose Rate (mrem/hr)

Exposure Time (hr) Number of Workers Frequency Dose (man-rem/yr) 1. Removal/replacement of insulation 150 80 4 1/yr 48 2. Installation/removal and ladders 50 40 4 1/yr 8 3. Inspecting inside drywell 150 80 6 1/yr 72 4. Recorder data 50 80 6 1/yr 24

5. Inspecting outside drywell 5 50 2 1/yr 0.5 Total 153 C OLUMBIA G ENERATING S TATION Amendment 53 F INAL S AFETY A NALYSIS R EPORT November 1998 12.4-16 Table 12.4-8 Occupational Dose Estimat e s During Special Maintenance

Activity Average Dose Rate (mrem/hr)

Exposure Time (hr)

Number of Workers Frequency Dose (man-rem/yr) Sparger replacement 800 60 5 Should not be nece s sary --- CRD replacement 260 35 5 1/yr 45.5 Turbine overhaul 5 250 20 1/5 yr 5 Servicing in-core det e ctors 15 50 3 1/yr 2.3 Offgas charcoal sys.

overhaul 20 100 2 1/20 yr 0.2 Special maintenance reactor water clearup sys.

150 100 8 1/10 yr 12 Misc. piping repairs 80 100 10 1/yr 80 Total 145.0 C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT A p ril 2000 LDC N-9 9-0 0 0 12.4-17 Table 12.4-9 Summary of Annual Information Reported by Commercial Boiling Water Reactors 1969 - 1980 Year Numb e r of Reactor s Included Annual Collective Doses (man-rems)

No. of Workers With Measurable Dose Gross MW-Yrs Electric Generated Average Dose Per Worker (rems) Average Collective Dose Per Reactor (man-rems)

Average No.

Person n el With Measurable Doses Per Reactor Average Man-rems Per MW-y r Average MW-y rs Generated Per Reactor Average Rated Capacity (MWe) Net 1969 3 (2) 586 (3 00) 290 a 192 1.03 a 195 145 a 3.1 64 112 1970 6 (4) 764 (5 10) 1,321 a 912 0.39 a 127 330 a 0.8 152 267 1971 7 (5) 1,784 (1 , 06 9) 1,873 a 1,038 0.57 a 255 375 a 1.4 187 339 1972 10 (7) 2,858 (2 , 13 0) 2,258 a 3,058 0.94 a 286 323 a 0.9 306 434 1973 12 4,564 5,340 3,394 0.85 380 445 1.3 283 459 1974 14 7,095 8,769 4,059 0.81 507 626 1.7 290 513 1975 18 12,611 14,607 5,789 0.86 701 812 2.2 321 611 1976 23 12,626 17,859 8,586 0.71 549 776 1.5 373 647 1977 23 19,042 21,388 9,098 0.89 828 930 2.1 396 645 1978 25 15,096 20,278 11,774 0.74 604 811 1.3 471 669 1979 25 18,322 25,245 11,671 0.73 733 1,010 1.6 467 669 1980 26 29,530 34,094 10,868 0.87 1,136 1,311 2.7 418 664 a During the y ears 1 9 69 t h rough 19 7 2, a ll plants rep o rted collecti v e doses b u t a few did not s ubmit the n u mber of per s onnel that re c eived measurable doses. The number of reac t ors that did report doses a nd nu m ber of work e r s is given in parentheses in the second c o l u m n. The collective doses shown in parentheses in the third column, as w e ll as the noted numb e rs in the r e maining columns, are all based on the data submitted by the number of reactors shown in parentheses. This c o rrection, and others, changed so m e of the values from those a ppearing in earlier NU R EG documen ts.

C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-11-029 12.5-1 12.5 RADIATION PROTECTION PROGRAM

12.5.1 ORGANIZATION

Radiation Protection, under the di rection of the Radiological Se rvices Manager, implements the Radiation Protection Program (RPP).

Health Physics (HP) is audited for compliance to regulations a nd to ensure that occupational and public radiation exposures are as low as is reasonabl y achievable (ALARA). Regulatory Guide 1.8 and ANSI 18.1-1971 have been followed in the selec tion of HP personnel. Energy Northwest pre-employment practices include screening to determine that plant employees are trustworthy, fit, and qua lified to perform their duties safely. The experience and qualifications of the personnel responsible for the RPP and for handling and monitoring radioactive materials including special nuclear source and byproduct materials, ar e described in Sections 12.5 and 13.1. Also, Section 13.1 describes the minimum qualificatio n requirements for specific plant personnel, using the criteria outlined in Regulatory Guide 1.8 and ANSI 18.1-1971.

The Plant General Manager reports to the Vice President, Operations and has the overall responsibility for the RPP. Th e Plant General Mana ger is responsible for ensuring that personnel, facilities, and other resources required to implement the RPP are avai lable. This includes ensuring that the authority to implemen t an effective RPP is delegated through the management structure, ensuri ng the program receives the active support of all Energy Northwest personnel, and ensu ring production goals, maintenance activities, and work schedules do not adversely affect the ability to provide proper ra diological controls. In turn, all plant personnel share the responsibility for en suring personal radiolog ical safety and are required to follow the rules and procedures established for ra diological safety. Specific responsibilities regarding ALARA are described in Section 12.1.

The Operations Manager reports to the Plant General Manage r and has the responsibility for ensuring the independence of the RPP from pl ant operational pressures. The Operations Manager provides the Radiological Services Manager the support necessary for the effective implementation of the RPP.

The Radiological Services Mana ger (RSM) reports to the Pl ant General Ma nager and is responsible for the implementation of the RPP. This position meets th e requirements of the Radiation Protection Manager (RPM). The RSM has direct access to the Plant General Manager in all matters relating to radiation safety. The RSM meets the qualifications defined in Regulatory Guide 1.8, and provides the experience and expertis e necessary to implement the RPP. RPM responsibilities may be assigned to any Radiati on Protection management or supervisory position desc ribed in this Chapter provided they meet the requirements of Regulatory Guide 1.8.

C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-13-039 12.5-2 The Radiological Services Manage r supports the safe, reliable, and economic operations of the plant within applicable laws, standards, codes, regulations, and Energy No rthwest policies.

The Radiological Support Supervis or reports to the Radiological Services Manage r and directs the activities and monitors the performance of the Radiological Planning Group and the Radiological Support Group. The Radiological Planning Group is responsible for performing ALARA reviews and evaluations to support HP Operations. Th e Radiological Support Group provides technical support fo r all aspects of the RPP.

The Radiological Operations Supervisor reports to the Radiological Services Manager and directs the activities and monitors the performance of the Health Physics Craft Supervisors and Rad Material Control/

Rad Waste Supervisor.

The Health Physics Craft Superv isors report to the Radiological Operations Supervisor and direct the activities and monitor the performance of HP Tech nicians. Health Physics Craft Supervisors are responsible for ensuring conditions that have th e potential for causing exposure to radiation are identified, posted, and controlled.

The Rad-Material Control and Rad-Waste Supervisor reports to the Radiological Operations Supervisor and is responsible for providing immediate supervisi on, leadership and technical support to the laborers in the areas of equi pment and area decontam ination, radioactive material control and inventory and anti-contamination laundry; pr ocess, package and transport of radioactive waste materi al, including mixed waste.

Each individual who has unescorted access to Colu mbia Generating Station restricted areas is responsible for ensuring personnel radiation safety. This in cludes strict compliance with radiation protection re quirements, procedures , and good radiological work practices. In addition, individuals with escort duties are responsible for the radiological safety of visitors.

12.5.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES This section describes the e quipment, instrumentation, and facilities available for implementation of the RPP and the criteria used for selection of the instrumentation and equipment. The guidance pr ovided by Regulatory Guides 8.

3, 8.4, 8.6, and 8.28 has generally been followed with exceptions noted as follows:

a. Regulatory Guide 8.3, "Film Badge Performance Criteria" will be followed if film badges are used in the plant progra m; however, other dosimeter types, such as optically stimulated luminescent (OSL) dosimeters, are used as the Dosimeter of Legal Record (DLR) for complian ce with 10 CFR 20.150 1, 20.1502, and 20.2206;

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 12.5-3 b. Regulatory Guide 8.4 is implemented fo r the selection of direct-reading pocket dosimeters as defined in S ection 2 of ANSI N13.5-197 2 except for C.2.b, which states, "The calibration/response te st result should not exceed +/-10% ofan exposure from a source traceable to the National Bureau of Standards."

• This is accepted on the minus side, but is considered excessively stringent on the positive side. Since the error on the positive side results in exposure conservatism to the worker, +20% is a mo re reasonable limit for rejection of a pocket dosimeter. Vendor literature will be accepted as documentation that performance standards specified in Regul atory Guide 8.4 are met. Continued use of direct-reading pocket dosimeters w ill be based on their ability to perform acceptably under test conditions for temperature and humidity described in approved Health Physics Inst ructions as follows: +/-2% drift per 24 hr at -10°C and any percent humidity; +2% drift per 24 hr at 50°C and 95% humidity; and

+20% and -10% of 80% of calibrated full scale;

c. Regulatory Guide 8.6, "Standard Test Procedure for Geig er-Mueller Counters," will be used as applicable. This guide references ANSI N42.3-1969 (ANSI/IEEE Standard 309-1970) for twelve different tests to Geiger-Mueller counters. Energy Northwest will develop tests and procedures to ensure that Geiger-Mueller tube characteristics are appropriate for planned (or intended) applications. These tests may incorporate plateau characteristics, dead time, efficiency, and operating environment;

d. The majority of direct-reading dosimet ers at Columbia Generating Station are electronic dosimeters with audible-alar m capabilities. A program for their appropriate use requires that conditions under whic h they may not perform adequately be discussed, as well as describing performance specifications which are met.

Electronic dosimeters will not be used to circumvent the initial meter survey required prior to work in an area. Ra diation Protection will assign electronic dosimeters only when the working envir onment is suitable for their use.

Individuals required to wear an electronic dosimeter will be provided appropriate instructions either in training or at the time of issue to minimize the risk of improper use. Regulatory Guid e 8.28 endorses, with one exception, the performance specifications indicated in ANSI N 13.27 "Performance Specifications for Pocket - Sized Alarming Dosimeters/Ratemeters."

• National Institute of Standards and Technology.

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 12.5-4 Since credit is taken for the audibl e-alarm capabilities of the electronic dosimeter, the Energy Northwest program for use of audible-alarm dosimeters complies with Regulatory Guide 8.

28 with the following exceptions:

1. Requirement: Section C.2.c of the Regulatory Guide requires a source check of audible-alarm dosimeters each day before use.

Energy Northwest position: Electro nic dosimeters equipped with an automatic electronic test to ensure de tector function are not subject to the requirement for a source ch eck each day before use.

2. Requirement: Section C.2.b(1) of the Regulatory Guide specifies that alarm dosimeters should not be used when the alarm may not be heard, such as (a) in a high noise envir onment, (b) when the user has a pronounced hearing loss, (c) when the user is wearing mufflers over the ears, or (d) when the sound from the dosimeter would be muffled by heavy clothing worn over the dosimeter.

Energy Northwest position: Electro nic dosimeters will be allowed for use when their alarm may not be h eard except when th ey are used to fulfill the alarming dosimeter f unction described in Technical Specification 5.7. When used in accordance with the Technical Specification, alte rnative methods of warning are required when the audible alarm may not be heard.

Alternative methods include, but are not limited to the following: (1) vibrator, (2) ear phone, (3) flashing

light clearly visible to the worker.

The program outlines the performance requirements for electronic dosimeters and details the exception to the AN SI N13.27 criteria, while ensu ring that reliable electronic dosimetry is used to facilitate expos ure control and the ALARA concept.

12.5.2.1 Criteria for Selection

a. Radiation and contamination survey instrumentation: This equipment was selected to cover the wide range requi rements extending from picocurie quantity measurements in the laboratory to th e thousand R/hour rang es necessary for emergency dose rate determinations. The laboratory instrumentation was chosen to provide capability for the quantitative and qualita tive analyses required to identify and measure the radionuclides encountered in a power reactor. The portable instrumentation includes low level detection capabilities for alpha, beta, and gamma contamina tion and wide rang es of dose rate measuring instruments for beta, gamma, a nd neutron radiation.

The criteria for quantity selection were to provide adequate available counting time for anticipated demand in the la boratory and sufficient porta ble instruments to cover C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-016 12.5-5 normal operational and emerge ncy requirements in all areas of the Columbia Generating Station facility;

b. Airborne radioactivity monitoring: The basic criteria for selection of this equipment were to provide a means fo r determining radioactive airborne effluents released from the plant, and to effectively monitor airborne radioactivity levels within the plant e nvirons. Provisions have been made for continuing response monitoring of noble gases discharged from gaseous release points from the reactor, radwaste, and turbine building, an d for continuous sampling of radioiodines and particulates at these same locations. Internal plant air monitoring instrumentation is used within these buildings with readout locally and in the control room;

c. Area radiation monitoring: This system was designed to provide continuous surveillance of radiation levels throughout the plant with local alarm at predetermined levels, local indication, and control room annunciation and recording. Functions of the system include warning of excessive gamma radiation levels in fuel st orage and handling areas, de tection of unauthorized or inadvertent movement of radioactive materials in the plant, local alarms to warn personnel in an area of a s ubstantial increase in radi ation levels, provision for supervisory information in the control room so that correct decisions may be made in the event of a radiation incide nt, backup to other systems for detection of abnormal migrations of radioactive materials in or from the process streams, and providing a permanent record of gamma radiation levels at selected locations within the various plant buildings; and

d. Personnel monitoring: Pe rsonnel dosimetry devices we re chosen to provide a record of exposure receive d by occupationally exposed individuals at the site who are likely to receive, under normal or accidental conditions, exposures greater than 10% of applicable 10 CFR 20 limits.

Personnel dosimeter badges (DLRs) cont aining an OSL dosimeter or other acceptable dosimetry provide the primar y legal record of exposure received by personnel. Each person requiring monitoring for record is assigned a badge, which is recorded with the wearer's identification.

Results from the badge and the period of exposure are recorded on a document kept as a legal Energy Northwest record. Badges used will be capable of recording exposure over a range of at least 40 mrem to greater than 1000 rem.

Persons being monitored may be required to wear other dosimetry assigned by the Radiation Protection staff, such as direct reading dosi meters, integrating dose meters, extremity ba dges, and finger rings.

C OLUMBIA G ENERATING S TATION Amendment 57 F INAL S AFETY A NALYSIS R EPORT December 2003 LDC N-0 1-0 6 9 12.5-6 12.5.2.2 Facilities

Radiation Protection facilities at Columbia Generating Stati o n include the following:

a. Personnel decontamination showers a nd sinks are located in the radwaste building (487 ft leve l). Te mporary change areas are s e t up as nece s s ary in areas of the plant to local i ze and prevent the spread of contamination while

performing maintenance activ i ties. Small inventories of protective clothing are

stored in the emergency reloca t ion ce n ters, operation and radwaste control rooms, and strategic locat i ons throughout the plant; b. Monitoring equipment f o r personnel [e.g., frisk e rs and installed personnel monitors (IPMs)] and tools

/personal items [fri s kers and small article m o nitors (SAMs)] are prov i ded at the radiological acce s s control areas and various areas

within the plant to survey f o r radio a ctive contamination; c. Facilities for personn e l exposure m onitoring and pr otection, which include:

1. Internal dosimetry,

2. Respiratory protection testing;

d. Medical fir s t aid faci l it i es a r e equ i pped to p r ov i de care for injuries, including those with radioactive c ontamination involved; and

e. Faci l it ies for equipment and tool d e contamination exis t in the radwaste, turbine, and reactor buildings. The locations and facil i t i es a r e

1. Radwaste building

The general decontamin a tion area is shown in F i gure 12.3-1 2 ,

approximate column location Q.4-13.

6 at the 467 ft 0 in. level.

Facilities include curbing, sink, monorail hoist, and drains. At the

487 ft 0 in. level, Figure 12.3-1 3 , column coo r dinates R.2-14.0, tools

and small equipment can be deconta m inated in the hot machine shop.

Facilities in the hot machine shop include a b e nch space and drains.

Also, there is a personnel (male/fem a le) decontamination station at the 487 ft 0 in. level, column coord i nate s K.1-15.9. This facility contains

sinks, showers, and a decontamination kit.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-056 12.5-7 2. Turbine building

Figure 12.3-5 , columns H-9.5, el. 441 ft 0 in. identifie s the turbine building decontamination area. Faciliti es include a monorail, curb, sink, shower, and drains.

3. Reactor building

The head washdown area is shown in Figure 12.3-18 at column coordinates N-5.8 at th e 606 ft 10 in. level and contains a curb and drain.

The CRD room area, Figure 12.3-16 , columns M-3.4, 501 ft 0 in.

elevation contains a sink, monora il, bench, and storage vault.

4. The office of the RPM is located in the service building.

Health Physics Craft Supervisors and HP Technicians are located in locations to provide for ready access by other plant work ers and an area to generate and process records.

5. A hot machine shop and a hot in strument shop are provided in the radwaste building for work on contaminated equipment under controlled conditions. A HEPA-filtered vacuum system is installed in the hot machine shop to control airborne radioactivity while working on radioactive equipment. Portable HEPA-filtered vacuum systems are also available.

6. A laboratory complex is provided in the radwaste building consisting of a sample room, hot radiochemistry laboratory, and a counting room where radioactive samples will be qualitativ ely and/or quantitatively analyzed.

f. A protective clothing storage and distribution facility inside the protected area fence, but outside the power block.

Radiation Protection facilities at the Plant Support Facility/Eme rgency Operations Facility:

The Energy Northwest Plant Support Facility (PSF) is located 0.75 mile southwest of Columbia Generating Station.

It is designed and e quipped to provide emer gency capabilities in support of Columbia Generating St ation and for support of Columb ia Generating St ation during normal plant operations and maintenance. Support facilities important to HP include:

a. Portable radiation mon itoring equipment calibration, b. Radiological training.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-05-056 12.5-8 The instrument calibration laboratory is located in the extreme northwest corner of the lower level. It contains an irradiation cubicle that is shielded on all sides and above by 2 ft of concrete. The cubicle entrance is protected by a labyrinth and a lockable gate. Larger open sources are stored and used in the cubicle. The shielded cubicle together with administrative controls such as procedures, radiation work permits, and surveys ensure that calibration laboratory operation will not result in radiation areas in surrounding spaces.

Calibrations are performed in accordance with a pproved procedures and are traceable, either directly or indirectly, to the National Institut e of Standards and Tec hnology (NIST). Available sources are listed in Table 12.2-12. 12.5.2.3 Equipment

Radiation Protection equipment, other than instrumentation, is described in the following:

a. Protective clothing and accessories are provided for personnel required to work in contaminated areas. Clothing requirements for a particular task or area are prescribed by Radiation Protection based on the actual or potential conditions.

Available clothing includes, but is not limited to:

1. Coveralls and laboratory coats, 2. Gloves - rubber and/or cotton, 3. Head covers, 4. Foot protection, and

5. Plastic suits - with or without supplied air.

b. Respiratory protection equipment is provided for personnel when it is not practicable to apply proce ss controls or other engin eering controls to control airborne radioactive contamination. The decision to use respiratory protection equipment is based on maintaining the TEDE ALARA. The respiratory protection program is conducted per the requirements of 10 CFR 20.1701, 1702, 1703, and 1704.

Exposure is limited to deri ved air concentrations (DAC) and annual limit on intake (ALI) values sp ecified in Appendix B, Table 1 of 10 CFR 20. Allowance is made for use of respiratory protective equipment, as prescribed in 20.1703, in limiting an individual's intake of airborne radioactive materials. Among the types of equipment used are:

1. Full face air purifying respirators, 2. Airline supplied full face masks (pressure demand regulated), and

3. Self contained breathi ng apparatus (pressure demand regulated).

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 LDCN-06-024 12.5-9 c. Air sampling equipment, in addition to the continuous air monitors, includes high and low volume portable air sample rs, low volume consta nt air samplers, and air samplers with a se lf-contained power source.

Collection media (filters) employed are capable of collecting par ticulate and radioiodine samples; and

d. Emergency equipment and supplies are maintained in lockers at strategic locations within the plant. These lock ers are to be used for a rapid initial response and are not intended to provide the resources for a long term recovery operation. Equipment is stored for field team use at the EOF. Four-wheel drive vehicles, automobiles, and survey kits are avai lable for use by the field team. Locations and types of emergency equipment are listed in the Emergency Plan. Other than emergency supplies, the primar y storage areas for radiation protection equipment are the two Radiation Protection control areas located in the service and radwaste buildings. Temporary storage fa cilities are set up in localized areas as required.

12.5.2.4 Instrumentation

Typical plant portable radiological instrumentation is described in Table 12.5-1. All of this instrumentation is calibrated at least semiannually when in us e except the Condenser R-meter which is calibrated annually. Electronic calibra tions of instrument co mponents are performed using test equipment traceable to the NIST. Overall calib ration of radiation measuring instruments is performed using radioactive standards traceable to a recognized source in a known, reproducible geometry.

Calibration of low level radia tion detection instruments is done with a pulse generator.

12.5.3 PROCEDURES

Section 12.1.3 described a process that was incorporat ed into the preparation and revision of plant procedures which provide a positive method of ensuring Radiati on Protection input and ALARA consideration into radiation exposure related activities. Th e intent of this process is to incorporate the general guidance of appr opriate regulatory guides plus the previous experience of power reactor radiation protection work into all applicable plant procedures.

12.5.3.1 Personnel Control Procedures

The Plant Procedures Manual contains the admini strative procedures for control of access to radiation areas, high radiation ar eas, and very high radiation ar eas. This includes control of time spent within these areas by all plant workers. Basica lly, the procedures limit entry to these areas to time required fo r necessary operational maintena nce and surveilla nce activities only. The primary tool used to ensure cont rol and to maintain TEDE ALARA at Columbia C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 12.5-10 Generating Station is the Radiation Work Permit (RWP). All work performed at Columbia Generating Station in ra diologically controlled areas is performed in accordance with an RWP, with the exception of specific activities identified in the implementing plant RWP procedure. The RWP provides current data on radiation levels within the area of interest, any restrictions on allowable work time, protective clothing and respiratory protective requirements, information on special tools or equipment needed, special radiation safety and personnel monitoring requirements, and any other special instructions or radiological hold points necessary. A section of the RWP is used to incorporate the criteria given in Regulatory Guides 8.2, 8.8, and 8.10 into each individua l task, even though it has already been included in job procedures through the system for ALARA consideration de scribed in Section 12.1.3. All RWPs require approval from Radiation Protection supervision prior to starting work. In addition, all personnel who perform activities covered by th e RWP are required to read, understand, and document their understanding as specified by implementing plant procedures. There are two types of RWPs:

a. Specific RWP is issued for the perfor mance of a particular task or function which falls outside the limitations imposed for General RWPs, and

b. General RWP may be issued to cove r repetitive (routine) functions in areas where radiological conditions are known and stable.

In addition to the administrative controls used at Columbia Generating Sta tion, certain physical controls are established which restrict entry to radiation ar eas, high radiation areas, and very high radiation areas. Radiati on areas are posted as required by 10 CFR 20, and high radiation areas and very high radiation areas are locked or otherwise controlled as specified by this same regulation and Technical Specifications.

The plant security control system complement s both the administrativ e and physical entry restraints by allowing access only to personnel with authorization to be in specific plant areas.

12.5.3.2 As Low As Is Reas onably Achievable Procedures

The procedures and processes described belo w are in addition to those described in Sections 12.1.3 and 12.5.3.1. They have been developed to ensure that o ccupational radiation exposures are maintained ALARA. A primary goal of the Columbia Generating Station RPP is the control and reduction of individual and collective radi ation exposures. This goal is achieved through training and comprehensive job planning and reviews as follows:

a. Training - ALARA training is required by plant procedures to be included in all applicable radiation training courses.

The training applies to individuals whose duties require working with radioactive materials, entering radiation areas, C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 12.5-11 directing the work of others in radiologically controlled areas, or planning work and preparing procedures for work in radiologically contro lled areas. ALARA training is commensurate with the individual's duties, responsibilities, and radiation exposure potential. Workers w ho may enter restricted areas are given specific instructions about prenatal exposure risks to the developing embryo and fetus. This instruction includes th e information provi ded by Regulatory Guide 8.13;

b. Job Planning and Reviews - Plant proce dures specify that each job, involving exposure to radiation and/or radioactive materials, receives job pl anning and/or an ALARA review. The extent of the re view is determined by an evaluation of the radiological risks involved. The preliminary job planning and review identifies the need for pr e-job briefings and/or j ob planning meetings to coordinate work efforts and to familiari ze personnel with th e work and exposure reduction techniques.

Pre-job planning may in clude the following:

1. Job history reviews, 2. Determination of radiological conditions, 3. Determination of exposure estimate, and

4. Interface with planners, schedulers, job supervisor, ALARA.

Additional ALARA reviews are performed by the Senior Site ALARA Committee. The Senior Site ALARA Co mmittee has been developed to ensure participation by a range of plant personnel and provide for an appropriate level of management invol vement and direction in ALAR A issues. The Senior Site ALARA Committee serves as a review a nd advisory organization to the Plant General Manager on occ upational radiation exposure to personnel.

Plant procedures provide require ments for committee membership, responsibilities, authority, and records of meetings and actions. The Senior Site ALARA Committee is responsible for the review of plant and departmental exposure goals and reviewi ng and assessing the effectiv eness of the radiation exposure control program and the ALARA Program. The Senior Site ALARA Committee may create Workin g Groups to provide dos e reduction methods for tasks which have a significant potential for dose reduction.

As part of pre-job revi ew, ALARA job planning m eetings are conducted when significant exposure savings or increased contamination control may result. The planning meetings may include the job supervisor, job planner, Radiation Protection supervision, HP technicians, and key workers. These meetings are used to ensure worker familiarity with procedures, work locations, RWP

requirements, unusual hazards, and jo b-specific ALARA techniques to be employed. The use of mock-ups or dry runs may result from these meetings.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 12.5-12 In addition to pre-job ALARA reviews, plant procedures have provisions for work-in-progress ALARA reviews and post-job ALARA reviews. These reviews are coordinated by Radiation Protection and may include discussions with individuals who performed the wo rk, HP technicians, engineers, job supervisors, designers, or others as appropriate;

c. Remote handling tools and/or equipmen t - use of special tools/equipment for remote handling of radioactive equipment is factored into each applicable work activity; and

d. Exposure records are maintained in a manner that will allow Radiation Protection to tabulate and correlate expos ure results to iden tify problem areas with individuals or activities.

Plant procedures are evaluated to determine the need for an ALARA review. An ALARA review is required for new procedures in whic h the actions take plac e in a radiologically controlled area or involve handling radioactive material.

An ALARA review is required fo r procedure revisions which:

a. Cause entry into a radiation area, hi gh radiation area, or high-high radiation area, b. Cause opening a contaminated or potentially contaminated system, c. Significantly increase dose rates, or

d. Significantly increase exposure received.

Radiation Protection requirements, prerequisites, precautions , and ALARA considerations are incorporated into these procedures during the procedure review and approval process. In addition, an RWP is issued for those activiti es having radiological implications. Special activities such as inservice inspection (ISI), outage, and refue ling are reviewed by the Senior Site ALARA Committee and Radiation Protecti on as appropriate. Special precautions, prerequisites, or requirements may be incorporated into the plant procedures based on these reviews.

12.5.3.3 Radiological Survey Procedures A radiological survey is defi ned as an evaluation of radiological c onditions and potential hazards. When appropriate, such an evalua tion includes a physical survey (e.g., direct radiation and/or cont amination surveys).

Routine surveys are conducted in various areas throughout the site to identify, monitor, and control sources of radiation and contamination. Routine surveys are performed on a frequency

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 12.5-13 based on a consideration of potential radiological hazar ds, personnel occupancy and radiological stability. Incl uded in routine surveys are the following daily checks:

• Check the plant area radiation monitoring system, and

• Check inservice portal, air and other continuous ly operating radiation/contamination monitors.

Abnormal changes in background and abrupt unexplained increases are investigated.

Nonroutine surveys are performe d as the need and conditions dictate. The frequency and extent of these surveys should be determined based on histori cal data, on the conditions and activities taking place in the area, and on ALARA considerations.

Surveys of normally inaccessible, unoccupied areas are performe d after each shutdown or prior to entry into these area

s. Postings and survey record sheets are updated as conditions dictate.

Instructions relating to radi ation surveys are provided in the Energy Northwest Radiation Protection Procedures.

12.5.3.4 Procedures for Radioactive Contamination Control

This section describes the bases and methods used for the monitoring and control of radioactive contamination on personnel, material, and surfaces.

a. Bases: The methods used for the monitoring and control of Columbia Generating Station licensed radioactive material are based on 10 CFR 20.1101(b), 20.1501, NRC Circul ar 81-07, and industry-accepted practices. Tools, equipment, and ot her items with dete cted quantities of licensed radioactive materials will not be unconditionally rele ased. Detection levels will be based on the ALARA principle.

b. Methods: Personnel a nd materials will be surveyed in accordance with 10 CFR 20.1501. When physical surv eys are performed, they will be conducted using industry-accep ted, calibrated, detecti on instruments and with techniques that are appropriate to the level of risk. Other sections of Chapter 12 cover the selection criteria for contamination survey instruments, contamination monitoring facilities, prot ective clothing, contamination and radiation controls established through the RWP program, contamination monitoring surveys, ALARA with respect to contamination, and control of airborne radioactive material.

C OLUMBIA G ENERATING S TATION Amendment 59 F INAL S AFETY A NALYSIS R EPORT December 2007 12.5-14 Instructions for monitoring and control of contamination are maintained in the Radiation Protection Procedures.

12.5.3.5 Procedures for Control of Airborne Radioactivity

Evaluation of airborne radioac tivity concentrations is done pr ocedurally by se veral methods.

Routine airborne surveys consist of observing the continuous air monitors located in various areas of the plant and al so the effluent monitors. These observations are supplemented by grab samples taken on a routine basis an d by laboratory analysis of selected particulate and charcoal filters used on the continuous monitors. Special airborne surveys are made with portable samplers when a continuous air monitor indicates increases in airborne radioactivity or to evaluate conditions in a specific area or on a specific job.

The portable air sampling equipment consists of both high and low volume collectors with appropriate media for collecting particulates and radioiodines. These samplers are used for both spot evaluations by collecti on of grab samples and longer te rm evaluations by use of low volume samplers to collect over the period of a specific job or activity.

Laboratory analysis is made of air samples for gro ss radioactivity and, where warranted, for specific isotope identification and quantification to determ ine and record airborne concentrations.

Selected numbers of the routine air samples collected are analyzed for specific isotope content to ensure that the DAC levels are not being approached. Special samples are taken for this purpose whenever unexplained increases occur on continuous air monitors or when gross activity levels indicate there is a potential for exceeding the value specified in 10 CFR 20, Appendix B, Table 1, Column 3 of a ny isotope present in the mixture.

Airborne radioactive iodine monitoring includes integrated sa mple collection and laboratory analysis plus portable sampling and analysis. Portable and stationary sampling encompasses iodine collection on charcoal and/or silver zeolit e cartridges of nominal di mension of 2-in. disc diameter by 1-in. thickness at calib rated flow rates. Duration of sampling is determined by the anticipated ambient concentration levels whereas a nominal sampling period in excess of 5 minutes is selected to mi nimize sampling errors. Where gross noble gas concentrations exist, the sample cartridge may be purged in the laboratory with clean filtered air to minimize noble gas interferences. The cartridge will be s ealed in a clean plastic bag and taken to the analytical laboratory counting room for analysis.

Areas are barricaded and posted as airborne radioactivity areas whenever average concentrations in that area exceed 0.3 DAC of the values specified in 10 CFR 20, Appendix B to Parts 20.1001-20.2401, Table 1, Column 3. The use of resp iratory protection equipment is evaluated when a significant potential for an airborne hazard exists, or when entering an area of unmonitored, unknown airborne contamination.

C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 LDCN-07-042 12.5-15 Various methods for control and reduction of airborne activity are incorporated into Energy Northwest Radiation Protection Procedures, wh ich include proper use of the ventilation system, use of specially designed equipment to collect radioactive airborne contaminants, methods for reducing and contai ning contamination to preven t it becoming airborne, and methods for cleanup of pr imary water prior to opening this system.

The respiratory protection program is designed to meet the requirements of 10 CFR 20.1701, 1702, 1703, and 1704.

Procedures for fitting, training, maintenance, and testing of the respiratory protection equipment are included. All equipment is required to have appropriate National Institute for Occupati onal Health and Safety (NIO SH)/Mine Safety and Health Administration (MSHA) approval if available. Unless the requirements are met, the protection factors are not used. Unapproved equipment ma y be used in some instances where reduction of intake of radioactive material will result, but no protection factor is taken for its use. An example of this is use of charcoal cartridges in atmospheres where radi oiodines are present to reduce the inhalation of these materials.

12.5.3.6 Radioactive Material Control Including Special Nuclear Materials (SNM)

Columbia Generating Station has implemented a program to ensure safe radioactive material control which include:

a. Procedures and training for receiving and shipping radioactive materials in accordance with 10 CFR 20.1906,

b. Procedures and training for storing licensed materials in accordance with 10 CFR 20.1801 and 1802,

c. Procedures and training for shielding, handling, and inventor y control of sealed and unsealed radioactiv e sources and SNM,

d. Procedures and training for posting and/or labeling radioactive materials in accordance with 10 CFR 20 requirements,

e. Procedures and training for leak testing sealed radioactive sources in accordance with Technical Specifications, and

f. Procedures and training for disposal of all licensed radioactive materials in accordance with 10 CFR 20, 10 CFR 30, 10 CFR 40, 10 CFR 61, or 10 CFR 70.

g. Procedures and training for activities associated with dry storage cask loading and unloading of spent fuel and the ope ration of the Independent Spent Fuel Storage Installation for storage of sp ent fuel in accordan ce with 10 CFR 72.

C OLUMBIA G ENERATING S TATION Amendment 62 F INAL S AFETY A NALYSIS R EPORT December 2013 LDCN-12-004 12.5-16 Inherent in the above mentioned procedures is direction for handling liquid standard solutions used for calibration of plant inst rumentation which include ventila tion control, shielding, waste collection, contamination control, and monitoring.

Plant procedures assign the responsibility for c ontrol and monitoring of sealed and unsealed sources and byproduct mate rials to the RPM. Th e Vice President Engin eering is responsible for overall implementation of control of SNM.

This is accomplis hed through a Nuclear Material Manager who is appoi nted in writing by the Vice Pr esident Engineering. The Chemistry Technical Supervisor is responsible for minimization of radioactive waste and the preparation, offsite ship ment, and disposal of radioactive materials and radwaste. Monitoring during handling of nucl ear materials is provided by Radia tion Protection, as appropriate.

12.5.3.7 Personnel Dosimetry Procedures

Section 12.5.2 describes the monitoring de vices used to provide the primary legal records of exposure incurred by pers onnel and additional equipment used to backup and supplement this data. Records of radiation exposure are maintained for each individual for whom personnel monitoring is required by 10 CFR 20.1502. Reports of required monitoring are documented on NRC Form 5 or electronic media containing a ll the information required by NRC Form 5. Energy Northwest provides these individual radiation exposure records pursuant to the provisions of 10 CFR 19.13.

For monitored individuals, a determination of prior occupa tional dose is made per the requirements of 10 CFR 20.2104. This includes the dose received during the current year at Columbia Generating Station and other nuclear facilities. This exposure history is documented on NRC Form 4 or equivalent.

All individuals who are monitored for external radiation exposure are monitored for internally deposited radioactivity as follows:

a. Initial, performed prior to the individual entering any radi ologically controlled area. Monitoring may be either quantitative (whole body count) or qualitative (passive monitoring).

b. When a worker formally declares pregna ncy, a whole body c ount is performed.

c. At termination of employment at Columb ia Generating Station, if the individual has been monitored for external radiati on exposure. Monitoring may be either quantitative (whole body count) or qualitative (passive monitoring).

C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-016 12.5-17 d. Whenever an individual causes an alarm of the passive whole body contamination monitors (portal monitors) and internal deposition is suspected, a quantitative bioassay (whole bod y count) is performed.

e. Special bioassays as de termined by the RPM.

Energy Northwest monitors all adult workers who are likely to exceed 10% of the 10 CFR 20 annual occupational radiation e xposure limits for adults. Radi ation exposure monitoring for minors (individuals less than 18 y ears of age) is required if they are likely to exceed 0.1 rem deep dose equivalent in a year. Monitoring for declared pregnant women is required if they are likely to exceed 0.1 rem deep dose equivale nt during the entire pre gnancy. Monitoring is not required for visitors who enter a restricted ar ea, since they are not lik ely to exceed 10% of the annual limit. However, confirmatory monitoring of visitors may be performed if directed by Radiation Protection s upervision. The determination of whether an individual is likely to exceed 10% of the 10 CFR 20 limit, and thus re quire monitoring, is based on a prospective evaluation. An evaluation is not required for each individual, but is based on employees with similar job functions.

For internal exposure, monitoring is required if an adult worker is likely to receive, in 1 year, an intake in excess of 10% of the applicable 10 CFR 20 annual limit. The need for internal exposure monitoring of i ndividuals is based on a prospectiv e evaluation whic h will be updated whenever there is an indication that there has been significant fuel failure.

Since 90 Sr and 3 H are not measurable by whole body counting, in-vitro bioassay (urinalysis) will be performed when the plant radiation surv eillance program indicates a potential need.

All results obtained from in-vivo and in-vitro bioa ssay will be evaluated a nd become part of the individual's record, as appropriate.

Energy Northwest complies with the adult occupational dose limits identified in 10 CFR 20.1201. An individual is allowed to exceed these 10 CFR 20 exposure limits only in exceptional situations wh ere the dose received is in accordance with the conditions of a planned special exposure, as specified in 10 CFR 20.1206. Records of planned special exposures are maintained and retained per th e requirements of 10 CFR 20.2105.

Written reports of planned special exposures are submitted to the NR C per the requirement s of 10 CFR 20.2204.

In addition to the 10 CFR 20 do se limits, Energy Northwest us es administrative exposure hold points to maintain exposures ALARA. Plant procedures allow an individual to exceed an administrative hold point but, only if a prio r approved dose exte nsion is obtained.

Procedurally, DLR badges are processed for radiation workers semiannually at a minimum but may receive interim processing if an abnormal exposure is suspected.

Pocket dosimeters and other auxiliary monitoring devices are used to maintain an estimate of an individual's dose during the interim period between processing of DLRs. The use of auxiliary monitoring C OLUMBIA G ENERATING S TATION Amendment 61 F INAL S AFETY A NALYSIS R EPORT December 2011 LDCN-11-016 12.5-18 devices as a permanent record of an individual's dose is restricted to times when DLRs are lost or damaged or give a false result. When a large discrepancy exists between the two devices, it must be established that the DL R is in error before the auxiliary monitoring device result is assigned as the permanent record.

Plant supervisors are notified of their assigned worker's exposure status and are responsible for maintaining these and their own exposure to ALARA and within specified limits.

Personnel dosimeters, that require processing to determine the radiation doses, are processed and evaluated in accordance with the requirements of the National Voluntary Laboratory Accreditation Program (NVLAP).

12.5.3.8 Radiation Protection Surveillance Program

The practices incorporated into the overall structure to ensure that the RPP is maintained at a high level and upgraded to meet new requi rements and problems are the following:

a. Section 12.1.1 describes the organization structured to provide assurance that the ALARA policy is effective. It is also pointed out in this section that the plant RPP has several levels of re view from a performance standpoint;

b. Section 12.1.3 describes the process for review of plant procedures for ALARA consideration;

c. The RWP program and other records pr eviously described provide a valuable source of information and are used to determine where the occupational radiation exposures are occu rring and as a means of re view for possible methods of exposure reduction;

d. The Radiological Servic es Manager and his staff work on an individual and group basis with other plant organizati ons to determine what their principal sources of exposure are and to look fo r methods of reducin g these exposures;

e. Procedures provide for routine mainte nance, calibration, and testing of all radiation instrumentation and equipment. New equi pment will be added as necessary for replacement and to supplement that existing. Written procedures are provided for use of equipment where required;

f. Plant facilities are routinely reviewed for possible improvements from a radiation protection standpoint. Section 12.1.3 describes several changes that have been incorporated into plant design for this purpose. Ot her considerations are additional shielding where practi cable, improved ventilation control, additional equipment, and incr eased physical restriction; C OLUMBIA G ENERATING S TATION Amendment 60 F INAL S AFETY A NALYSIS R EPORT December 2009 12.5-19 g. The routine and special surveys previously descri bed point out levels of radioactive contamination in plant areas.

The Columbia Generating Station staff is committed to maintaining a clean plan t and considers it routine procedure to reduce levels of contamination whenever such action will not result in an increase of occupational radiation exposure to personnel;

h. One aspect that is considered important and used in implementing the RPP is the incorporation of previous reactor and pow er reactor experience in this area. Previously successful methods, procedur es, and equipment are used whenever possible; and

i. Training of all personnel w ho work in the plant in radi ation safety practices is mandatory and given a high priority by Energy Northwest and Columbia Generating Station Ma nagement. The Training Mana ger, in conjunction with the Radiological Services Manager, is responsible for development of all

training programs, including radiation safety indoctrination.

Radiation Protection assists in this training by providing instructors for some phases. The degree of training provided each plant worker is dependent on his function and degree of responsibility; however , all radiation work ers in the plant are provided training considered necessary or required for their position. The training programs provided are desi gned to meet the requirements of 10 CFR 19.12 and the guidance of Regulator y Guides 1.8, 8.8, 8.10, 8.27, and 8.29. Clarifications, ela borations, and exceptions in using the above mentioned regulatory guides are located in the Energy Northwest Procedures.

C OLUMBIA G ENERATING S TATION Amendment 54 F INAL S AFETY A NALYSIS R EPORT April 2000 LDC N-9 9-0 0 0 12.5-21 Table 12.5-1 Health Physics Instrumentation Type Number Available Radiation Detected Sensitivity Range Ion chamber dose rate survey

meter 9 Beta, gamma 0-5E4 mR/hr (gamma) 0-2.E4 mrad/hr (beta)

cfx 5R/hr High range ion chamber dose

rate survey meter 2 Gamma 0-1.999E7 mR/hr Telescoping dose rate survey

meter 3 Gamma 0-1.0E3 R/hr Neutron do s e rate sur v ey meter 2 Neutron 0.1-5.0E3 mrem/hr Contamination survey meter with end window or pancake GM

probe 20 Beta, gamma 0-5.0E4 cpm or 0-5.0E5 cpm Contamination survey meter with

scintillation detector 2 Alpha 0-5.0E5 cpm 0 Condenser R-meter 1 Gamma 0-100 R Direct reading pocket dosimeters 300 Gamma 0-999 rem Direct reading pocket dosimeters 200 Gamma Various ranges 0-500 m R