ML18016A413

Rev 10 to ODCM for Shearon Harris Nuclear Power Plant.

ML18016A413
Person / Time
Site:	Harris
Issue date:	12/18/1997
From:	CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML18016A410	List: ML18016A409 ML18016A412 ML18016A413
References
PROC-971218, NUDOCS 9805060079
Download: ML18016A413 (453)

Text

SHEARON HARRIS NUCLEAR POWER PLANT OFF-SITE DOSE CALCULATION MANUAL (ODCM)

Revision 10 Docket No. STN-50-400 CAROLINA POWER & LIGHT COMPANY Approval by PNSC Chairman Approval by General Manager - Harris Plant Effective Date RECT(QFg DEC 1 8[)@,

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shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE OF CONTENTS Section Title Pacae TABLE OF CONTENTS LIST OF TABLES LIST OF FIGURES . . ~ ~

LIST OF EFFECTIVE PAGES v3.

1.0 INTRODUCTION

2.0 LIQUID EFFLUENTS 2-1 2.1 Compliance With 10 CFR 20 2-2 2.2 Compliance With 10 CFR 50 Appendix I 2-13 3.0 GASEOUS EFFLUENTS 3.1 Monitor Alarm Setpoint Determination 3-1 3.2 Postrelease Compliance With 10CFR20-Based ODCM Operational Requirement 3.11.2 3-15 3.3 Compliance With 10 CFR 50 3-22 4.0 RADIOLOGXCAL ENVIRONMENTAL MONXTORXNG PROGRAM 5.0 INTERLABORATORY COMPARISON STUDIES - .

6.0 'TOTAL, DOSE (COMPLIANCE WITH 40 CFR 190) 6.1 Dose to the Likely Most Exposed Member of the Public 6.2 Dose to a Member of the Public Due to Activities Within the Site Boundary 6-2 7.0 LXCENSEE-INITIATED CHANGES TO THE ODCM APPENDIX A METEOROLOGICA DXSPERSION FACTOR COMPUTATIONS APPENDIX B DOSE PARAMETERS FOR RADIOIODINES~ PARTICULATES/

AND TRITIUM APPENDIX C RADXOACTXVE LXQUID AND GASEOUS EFFLUENT MONITORING INSTRUMENTATXON NUMBERS APPENDIX D - PROGRAMMATIC CONTROLS D-1 APPENDIX E PROGRAMMATXC CONTROL BASES APPENDIX F ADMINXSTRATXVE CONTROLS APPENDIX G DEFINITIONS

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 LIST OF TABLES No. Title Pacae 2.1-1 Liquid Effluent Release Tanks and Pumps 2-19 2.1-2 Setpoints for Cooling Tower Blowdown Dilution Flow Rates (Fgygn) 2-20 2.1-3 Signal Processor Time Constants (x) for GA Technologies RD-53 Liquid Effluent Monitors ~ ~ ~ ~ ~ ~ ~ ~ 2 20 2.1-4 Nuclide Parameters 2-21 2.2-1 Q, Values .for the Adult for the Shearon Harris Nuclear Power Plant . . . . . . . . . . . . . . . . . . . . . . . . 2-24 3.1-1 Gaseous Source Terms 3-12 3.1-2 Dose Factors and Constants ~ ~ 3 13 3.1-3 Gaseous Monitor Parameters 3-14 3.2-1 Releases from Sheazon Harris Nuclear Power Plant (1)

Normal Operation (Curies/year) 3-18 3.2-2 Distance to the Nearest Special Locations for the Shearon Harris Nucleaz Power Plant (Miles) 3-19 3.2-3 Dose Factors for Noble Gases 3-20 3.2-4 Pq, Values (Inhalation) for a Child 3-21 3.3-1 through 3.3-19 R Values for the Shearon Harris Nuclear Power Plant ~ 3-34 F 1 Radiological Environmental Monitoring Program ~ ~ ~ 4 2 A-1 through A-4 X/Q and D/Q values for long-term gzound-level releases at special locations A-4

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 LIST OF TABLES (continued)

No. Title Pa<ac A-5 through A-6 Undepleted, No Decay, X/Q Values For Long Term Ground Level Releases At Standard Distances (Sec/M') A-5 A-7 through A-8 Undepleted, 2.26 Day Decay, X/Q Values For Long Term Ground Level Releases At Standard Distances (Sec/M') A-7 A-9 through A-10 Depleted, 8.0 Day Decay, X/Q Values For Long Term Ground Level Releases At Standard Distances (Sec/M') . . . . '. . . . A-9 A-11 through A-12 A-13 Distances Joint (m ............. ~.........

Deposition Values (D/Q) For Long Term Releases At Standard

')

Wind Frequency Distribution By Pasquill Stability A-11 Classes At SHNPP . . . . . . . . . . . . . . . . . . . . . . A-13 A-14 Shearon Harris Plant Site Input Information for Continuous Ground-level Release Calculations Wi,th the NRC XOQDOQ Program A-16 B-1 Parameters for Cow and Goat Milk Pathways B-16 B-2 Parameters for the Meat Pathway . . . . . . . . . . . . . . . B-17 B-3 Parameters for the Vegetable Pathway . . . . . . . . . . . . B-18

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 LIST OF TABLES (continued)

No. Title Pacae NOTE: Tables in Appendix D are named after their respective Operational Requirement.

3.3-12 Radioactive Liquid Effluent Monitoring Instrumentation D-3 4.3-8 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements D-5 3.3-13 Radioactive Gaseous Effluent Monitoring Instrumentation D-8

4. 3-9. Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements D-10 4.11-1 Radioactive Liquid Waste Sampling and Analysis Program D-13 4.11-2 Radioactive Gaseous Waste Sampling and Analysis Program D-19 3.12-1 Radiological Environmental Monitoring Program . D-29 3.12-2 Reporting Levels for Radioactivity Concentrations in Environmental samples D-35 4.12-1 Detection Capabilities for Environrqental Sample Analysis Lower Limit of Detection (LLD) D-36 Frequency Notation ~ ~ ~ ~ ~ ~ G 4 G-2 Operational Mode . . . . . . . . . . . . . . . . . . . ~ . . G-5

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 LXST OF FXGURES No Title ~Pa e 2.1-1 Liquid Waste Processing Flow Diagram 0 2 27 2 '-2 Liquid Effluent Flow Stream Diagram . 2-28 2~1 3 Normal Service Water Flow Diagram . 2-29

2. 1-4 Other Liquid Effluent Pathways 2-30 3.1 SHNPP Gaseous Waste Streams 3-53 3.2 Schematic of Airborne Effluent Release Points 3-54 3.3 SHNPP Condenser Off-Gas Syst: em 3-55

4. 1-1 SHNPP Exclusion Boundary Plan 4-16

4. 1-2 through 4.1-5 Environmental Radiological Sampling Points 4-17

Shearon Harris Nuclear Power Plant (SHNPP) December, 1997 Offsite Dose Calculation Manual (ODCM) Rev. 10 List of Effective Pa es Pacae Revision Cover Page 10 ll-lli lv Vl 10 Vl1 10 6

2 2-8 6 2-9 2-10 2-11 6 2-12 7 2 2-14 4 2-15 6 2-16 ~ I 2-17 2-19 6 2-20 4 2-21 2-23 10 2-24 27 4 2-28 6 2-29 4 2-30 6 3-1 6 3-2 5 3-12 8 3-13 5 3-15 3-17 6 3-18 8 3-19 3-23 5 3-24 6 3-25 3-26 5 3-27 3-28 6 3-29 3-30 5 3-31 3-33 6 3-34 3-52 5 3-53 6 3-54 3-55 5 Vi

Shearon Harris Nuclear Power Plant (SHNPP) December, 1997 .

Offsite Dose Calculation Manual (ODCM) Rev. 10 List of Effective Pa es Pacae Revision 4-1 4 4-3 6 4 4-7 4 4-8 io 4 4-10 4 4-11 10 4-12 4 4-13 6 4-14 .o )

4 4-18 4 4-19 4-20 1 5-1 6 6 6-2 6 7-1 6 A-1 4 A-2 6 A-3 4 A A-21 6 B"1 6 B-2 5 B-3 6 B-4 5 B B-18 4 C-1 D D-2 6 D-3 9 D-4 6 D-5 7 D-6 .o D D-10 6 D-11 9 D-12 D-40 6 E E-12 6 F F-5 6 G G-5 6

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 1.0 ZNTRODUCTZON The Off-Site Dose Calculation Manual (ODCM) provides the information and methodologies to be used by Shearon Harris Nuclear Power Plant (SHNPP) to ensure compliance with Operational Requirements 3.3.3.10, 3.3.3.11, 3/4.11.1I 3/4.11.2, 3/4.11.4, 4.12.1, 4.12.2, and 4.12.3 and reporting requirements in Appendix F of the ODCM. These operational requirements l are those related to normal liquid and gaseous radiological effluents, environmental monitoring, and reporting. They are intended to show compliance with 10CFR20-based requirements, 10CFR50.36a, Appendix Z of 10CFR50, and 40CFR190 in terms of appropriate monitoring instrumentation, setpoints, dose rate, and cumulative dose limitations.

Off-site dose estimates from non-routine releases will be included in the cumulative dose estimates for the plant to comply with Appendix Z of 10CFR50.

The ODCM is based on "Westinghouse Standard Technical Specifications" (NUREG 0452), "Preparation of Radiological Effluent Technical Specifications foz Nuclear Power Plants" (NUREG 0133), and guidance from the United States Nuclear Regulatory Commission (NRC). Specific plant procedures for implementation of this manual are presented in the SHNPP Plant Operating Manual.

The ODCM has been prepared as generically as possible in order to minimize the need for future revisions. However, some changes to the ODCM are expected in the future. Any such changes will be properly reviewed and approved as indicated in Administrative Controls Section 6.14 of the SHNPP Technical Specifications.

The assessment of annual radiation doses to members of the public from radibactive liquid and gaseous effluents from the plant is estimated using the NRC codes LADTAP ZZ -and GASPAR and concurrent meteorology for the report period. These off-site dose estimates for each calendar year are reported in the Annual Radioactive Effluent Release Report required by Appendix F of the ODCM.

1-1

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 2.0 LIQUID EFFLUENTS Radioactive materials released in liquid effluents from SHNPP to unrestricted areas are required to demonstrate compliance with 10 CFR 50 Appendix I (ODCM Operational Requirement 3.11.1.2) and, on an annual average basis, be limited to the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2. For dissolved or entrained noble gases the concentration shall be limited to 2E-4 pCi/ml total activity. On an individual release basis, the release concentration for liquid effluents will be limited to ten times (10x) the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2 (ODCM Operational Requirement 3.11.1.1). The liquid effluent release point is at the point of dischazge from the Cooling Tower Blowdown Line into Harris Lake (see Figure 2.1-3 and T/S Figure 5.1-3).

Figure 2.1-1, Liquid Waste Processing Flow Diagram, and Figure 2.1-2 Liquid Effluent Flow Stream Diagram, show how effluents aze processed and where they aze released.

Effluent monitor identification numbers are provided in Appendix C.

Liquid effluent dilution prior to release to Harris Lake is provided by the Cooling Tower Blowdown Line. Concurrent batch releases shall not occur at SHNPP.

The Secondary Waste Sample Tank (SWST) and the Normal Service Water (NSW) system have a low potential for radioactive effluent releases.

These releases are checked by effluent monitors on the SWST (Figure 2.1-

2) and the NSW lines (Figure 2.1-3).

The Turbine Building floor drains and the outside tank area drains (Figure 2.1-4) are monitore'd effluent lines with low probability of radioactive contamination.

The radioactive liquid waste sampling and analysis required for batch and continuous releases are found in Table 4.11-1 of the ODCM Operational Requirements.

The SHNPP ODCM uses the Canberra, Inc., Effluent Management System (EMS) software for automating the necessary calculations and record keeping.

As such, the ODCM is written with the following parameters set in the EMS:

The SET OPT option is set to NO WASTE The SETP EQN option is set to LOW ACT 2"1

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 2.1 Com liance with 10 CFR 20 10 CFR 20.1301 requires that the total effective dose equivalent to individual members of the public will not exceed 0. 1 rem (100 mrem) in a year.

10 CFR 20.1302 states that a licensee can show compliance with the annual dose limit of 20.1301 by demonstrating that the annual average concentration of radioactive material released in liquid effluents at the boundary of the unzestricted area does not exceed the values specified in 10 CFR 20, Appendix B, Table 2, Column 2.

ODCM Operational Requirement 3.11.1.1 states that, on an individual  !

release basis, the concentration of radioactive material released in liquid effluents to unrestricted area shall be limited to 10 times the values specified in 10 CFR 20, Appendix B, Table 2, Column 2.

ODCM Operational Requirement 3.3.10 requires that radioactive effluent instrumentation have alarm/trip setpoints that will ensure that an alarm/trip will occur prior to exceeding 10 times the limits of ODCM Operational Requirement 3.11.1.1. for principal gamma emitters. I Liquid effluent monitors have two setpoints, the high alarm and the alert alarm. The high alarm setpoint, S , provides alarm and isolation if the radionuclide concentrations, when diluted, would approach the ODCM Opezational Requirement limits for concentrations in unrestricted areas. Alert alarm setpoints, S,~ are set at a fraction of the S to provide an early warning of the approach .to ODCM Operational Requirement limits.

2-2

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 2.1.1 Batch Releases Radioactive liquids are routinely released as batches from the Waste Evaporator Condensate Tank (WECT) and Treated Laundry and Hot shower Tank (TL&HST) . Batch releases may also originate from the Secondary Waste Sample Tank (SWST) and Waste Monitor Tank (WMT). These tanks are shown in Figures 2.1-1 and 2.1-2. Based on analysis of the tank contents, the tank release rate is adjusted, based on the Cooling Tower Blowdown,Line flow rate, to dilute the tank activities to 50 percent of the allowable concentrations at the release point to Harris Lake.

The ODCM software calculates a nuclide specific response setpoint which is based on the sum of responses for each nuclide. The nuclide specific response setpoint equates all gamma-emitting nuclides to Cs-137, to which the monitor is calibrated.

If analysis of the batch sample indicates all gamma-emitting nuclides

. are < LLD, (as dhfined in ODCM Operational Requirement Table 4.11-1),

the tank gamma activity, Cz, may be assumed to consist only of Cs-134.

This nuclide has the lowest Effluent Concentration Limit (ECL) of any to be found in liquid effluents and provides a conservative basis for a monitor setpoint.

2-3

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 2.1.1 Batch Releases (continued)

Minimum Tank Mixing Time Footnote 2 to ODCM Operational Requirement Table 4.11-1 requires that the method used to mix an isolated effluent tank prior to sampling and analysis be described.

Equation 2.1-1 below provides an acceptable method for ensuring a well mixed tank so that a representative sample can be taken for radioactivity or other appropriate analyses.

(V) (E) (n)

(2. 1-1)

(RR) (60) where:

R = Minimum allowable mixing time, hr V = Tank capacity, gal E = Eductor factor RR Pump design recirculation flow rate, gpm n = Number of tank volumes for turnover; this will be a minimum of two

~

60 60 min/hr Table 2.1-1 lists the tarik capacities, eductor factors, and pump design recirculation flow rates for individual liquid effluent release tanks.

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 2.1.1 Batch Releases (continued)

2. Required Dilution Factor ODCM Operational Requirement 3.11.1.1 requires that the sum of concentrations divided by ECL values must not exceed 10 for an individual release. Therefore:

Z(~ ECL~

5 10 (2. 1-2) where:

C~ = the concentratinn of nuclide i to be released ECL< = the Effluent Concentration Limit for nuclide 10CFR20, Appendix B, Table 2, Column 2.

i from If the summation is greater total required dilution than 10, dilution is required. The is the minimum acceptable factor, D,~,

dilution factor required to meet the limits of ODCM Operational Requirement 3.11.1.1, based on pre-release and composite analysis.

Dreq re, y + D xeq, ng (2. 1-3) where:

X<<f g Required dilution factor for gamma-emitters ECL (2.1-4) f ~ R 2-5

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 2.1.1 Batch Releases (continued)

D~,~ = Required dilution factor for non-gamma-emitters g.ag ECL)

+

r (2. 1-5) f ~ R 5 ~

Q and f = 0.5 A safety factor to assure that the nuclide concentrations are 50% of the ODCM Operational Requirement limit at the point of discharge.

a value to take into account that tritium is potentially being released via the settling basin discharge to the cooling tower discharge line. This value is normally set to 1E-03, which is the H-3 ECL.

R = The maximum ECL ratio for the release point (normally set to 10) .

The sums include gamma-emitters (g) and non-gamma-emitters (ng),

respectively.

The measured concentration of each gamma-emitting nuclide,

~ including noble gases, is reported in pCi/ml. If no gamma activity is detectable then an activity of 9E-07 pCi/ml of Cs-134 is assumed for setpoint calculations. The measured concentration of non-gamma emitters is determined by analysis of the liquid effluent or previous composite sample, and is reported in pCi/ml.

2-6

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 2.1.1 Batch Releases (continued)

3. Maximum Waste Flow For liquid releases, the maximum permissible waste flow rate for this release, W~ is the minimum of R and R where avail alloc R~x (2. 1-6) teq R = Liquid effluent tank maximum waste flow rate, as specified in Table 2.1-1. This value is the same as F~.t. ~

and F avail The available dilution flow is the minimum dilution stream flow (Cooling Tower Blowdown) that can be ensured for the period of the release. Since only one batch release occurs at a time out of a single discharge point, the flow is not corrected for other releases in progress, for any activity in the dilution stream, or reduced by a safety factor. The minimum dilution flow rate for each setting is shown in Table 2.1-2.

falloo Fraction of the available dilution volume which may be assigned to a particular release to ensure. discharge point limits are not exceeded by simultaneous radioactive liquid releases. The value of F lloo is based on assumed operational considerations for simultaneous releases but normally will be 0.8 for a batch release and 0.2 for a continuous release.

Minimum Dilution Fldw Rate The Minimum Dilution Flow Rate (min dflow) is the minimum Cooling Tower discharge flow necessary to dilute the release to less than ODCM Operational Requirement Limits.

If Dq s 1, the minimum dilution flow rate is set to 0.0. If Dq

> 1, the minimum dilution flow rate is determined as follows:

~ D teq min dflow waste (2.1-7) a 1 loc where F , = waste flow anticipated for this release 2-7

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 2.1.1 Batch Releases (continued)

5. Setpoint Calculations The ODCM software calculates a nuclide specific response setpoint, which is based on the sum of responses for each nuclide. The setpoint equates all gamma-emitting nuclides to Cs-137, to which the monitor is calibrated. The setpoint is listed in terms of Cs-equiv and the units are pCi/ml.

If analysis of the batch sample indicates all gamma-emitting nuclides are < LLD, (as defined in ODCM Operational Requirement Table 4.11-1), the tank gamma activity, C~, may be assumed to consist only of Cs-134. This nuclide has the lowest ECL of any to be found in liquid effluents and provides a conservative basis for a monitor setpoint.

(1) Maxi'mum setpoint value, based on Nuclide Specific Response S~ (Cs-equiv) (Sq 'R ) +B (2. 1-8) where Setpoint ad)ustment factor.

alloc avail FMRS'tO roy,ny (2. 1-9) req,g S,~> should always be greater than 1 to ensure that adequate dilution flow is available for the release.

B = monitor background (pCi/ml) 2-8

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 2.1.1 Batch Releases (continued)

R, = g slope~ C~

where the sum extends over all nuclides which have response factors stored in the database for the'onitor of interest and slope< = the Liquid Effluent Monitor Gamma Sensitivities (from Table 2.1-4) for nuclide i, relative to Cs-137. To make nuclide i relative to Cs-137, the nuclide sensitivity is divided by the Cs-137 sensitivity.

Sensitivit (nuclide i)

Cs-137 Sensitivity (2) Monitor alert alarm setpoint, S~, (Cs-equiv)

An Alert Alarm setpoint is calculated to provide an operator with adequate warning that the high alarm setpoint is being approached. S, is calculated from the nuclide specific response setpoint.

(2. 1-10) where:

A value (1.0 designed to provide an operator with adequate warning that the high alarm setpoint is being approached.

2-9

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 2.1.1 Batch Releases (continued)

(3) Check for Excessive Monitor Background In order to differentiate between the S, and the statistical fluctuations associated with a high monitor background, a check for excessive monitor background is made. As a check, verify that the minimum detectable concentration (MDC) for the monitor is less than 0 ' of the net S therefore, background is acceptable if:

O.lf (S - B)'] (2. 1-11) where:

MDC (2.1-12) where:

Signal Processor Time constant, minutes.

(Table 2.1-3)

Bkg = Background- Count Rate, in cpm B/ ER Monitor efficiency for the Cs-137 gamma energy, cpm/pCi/ml determined by primary calibration.

If not, postpone the release and decontaminate or replace the sample chamber to reduce the background, then recalculate S and S, using the new, lower background.

Post-Release Compliance After the release is made, actual concentrations are used to check 10 CFR 20 limits, and the actual dilution flow and waste flow are used instead of the anticipated dilution flow and waste flow.

For batch releases, the duration is determined from the start and end dates and times of the release. This is used with the actual release volume to calculate the release rate.

l 2-10

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 2.1.2 Continuous Releases The continuous releases from the SWST and the NSW return lines 'are monitored as shown in Figures 2.1-2 and 2.1-3. The function of these monitors, in contrast to the isolation function of batch release tank monitors, is to provide an indication of low levels of radioactivity in the effluent. The continuous effluent monitor setpoint is based on an assumed FSAR nuclide mix for the SWST (from Table 11.2.1-5 of the FSAR).

The EMS software does not calculate continuous release monitor setpoints.

1. Monitor High Alarm Setpoint, S (pCi/ml).

0.1 (ECL n Sens rr ) + g S~x (2.1-13) where:

EC Lore Weighted EC for the SWST nuclides listed in Table 11.2.1-5 of the FSAR.

Sens,ff Z> (Sens i x % abundance) nu.x, cpm/pci/ml.

for the SWST nuclide Monitor Alert Alarm Setpoint, S,~~ (Cs-equiv)

( (S -'B)'] + B (2.1-14)

When the monitor is operable and not in alarm, analysis of weekly composite samples is not required by ODCM Operational Requirement Table 4.11-1.

If the monitor is in alarm or the presence of non-naturally occurring radioactivity > effluent LLD is confirmed, the releases may continue provided the sampling and analysis required by ODCM Operational Requirement Table 4.11-1 are performed. The results of the sample analysis will be evaluated for compliance with ODCM Operational Requirement 3.11.1.1.

The monitor alarm setpoints may be recalculated using the methodology in Section 2.1.1 with the results of the gamma analysis and analyses of the composite sample.

3. Check for Excessive Monitor Background Monitor background is considered excessive when the minimum detectable concentration (MDC) for the monitor is >0.01 ECL,fg.

Therefore, background is acceptable if:

0 01

~ (ECL ~~

'ENS ff )

MDC (2. 1-15) 2-11

shearon Harris Nuclear Power Plant November 1995 Offsite Dose Calculation Manual (ODCM) Rev. 7 2.1.3 Ot ez Li u'd Releases

1. Outdoor Tank Area Drain Effluent Line The outdoor tank area drain effluent line routes rain water collected in the outdoor tank area to the storm drain system and from there directly to the lake. The line is monitored for radioactivity by the Tank Area Drain Transfer Pump Monitor.

Because no radioactivity is normally expected in this line, the monitor high alarm and alert alarm setpoints are determined using the methodology is Section 2.1.2. lf the setpoint is exceeded, the discharge pump is automatically secured. Effluent can then be diverted to the floor drain system for processing and eventual release (see Figures 2.1-1 and 2.1-2).

Turbine Building Floor Drains Effluent Line Water collected in the turbine building floor drains is normally routed to the yazd oil separator for release to the environment via the waste neutralization system and then to the cooling towe discharge line. Tritium is expected to be detected in this pathway fzom sources such as background from the lake. Because nc other radioactivity is normally expected in this path, the setpoints for the tuzbine building drain monitor are determined using the methodology in Section 2.1.2. Should the setpoint be exceeded, the release is automatically terminated. Effluent can then be diverted to the secondary waste treatment system for processing and eventual release (see Figures 2.1-1 and 2.1-2).

\

2-12

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 2.2 Com liance with 10 CFR 50 A endix I 2.2.1 Cumulation of Doses The dose contribution from each release of liquid effluents will be calculated and a cumulative summation of the total body and each organ dose will be maintained fo" each 31 days (monthly), each calendar quarter, and the year.

The EMS software calculates and stores the dose for the critical receptor (site boundary), for each nuclide, and for each organ. The dose is the total over all pathways which apply to that receptor. A receptor is defined by receptor ID, age group (infant, child, teen, or adult), sector, and distance from the plant.

The dose contribution for batch releases and all defined periods of continuous release received by receptor "r" from a released nuclide "i" will be calculated using the following equation:

Distr Al 'Eht C),F (2. 2-1) where:

Dl,: "

the cumulative dose or dose commitment to the total body or an organ "x" by nuclide "i" for receptor "r" from the liquid effluents for the total time period of the release, in mrem.

A,I IC site-related ingestion dose or dose commitment factor fo" receptor "r" to 'the total body or organ "x" for nuclide in mrem/hr per pCi/ml.

length of time period 's', over which the concentration and F value are averaged, for all liquid releases, in hours.

C i% the average concentration of nuclide "i" in undiluted liquid effluent during time period at, from any liquio release, pCi/ml .

the near field average dilution factor for receptor "r" during any liquid effluent release.

2-13

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 2.2.1 Cumulation of Doses (continued)

Where:

F waste Fss (2.2-2) waste avail and mixing ratio fraction of the release that reaches the receptor. At the SHNPP, this value is set to 1.

Also, the sum extends over all time periods 's'.

Zn the case of a continuous secondary'aste sample tank radioactive release, Cz = the concentration of nuclide "i" in the SWST composite sample. For the NSW, Cz = concentration of nuclide "i" in the cooling tower basin and F = discharge from the cooling tower basin while F,,n = the flow from the makeup water cross-tie. For a release through the Turbine Building Floor Drain Line to the waste neutralization system, Cq = the Turbine Building floor drain sample activity, F~~ = discharge from the Turbine Building floor'rain line, and F the avexage flow during the period of the total Cooling Tower discharge.

The total Cooling Tower discharge is the,sum of the Cooling Tower Blowdown flow and the Cooling Tower Bypass Line flow.

When there is a primary-to-secondary leak, the change in concentration of tritium in the steam generators times the secondary loses (balance of plant), will be used for effluent accountability. The secondary loss rate will also be used for volume accountability.

The dose factor g, (see NURE6-0133, Section 4.3.1) was calculated for an adult for each isotope "i" using the following equation:

1.14 +05( + 21BFg) DFf'w (2. 2-3) where:

The ingestion dose commitment factor to the whole body or any organ "xw fox an adult for each nuclide "i" ~

Corresponding to fish consumption from the Harris Lake (dilution = 1) and drinking water from Lillington (dilution = 13.95).

Values for the adult total body and organs in mrem/hr per pCi/ml are given in Table 2.2-1.

2-14

Shearon Harris Nuclear Power- Plant (SHNPP) August 1995 Of fsite Dose Calculation Manual (ODCM) Rev. 6 2.2.1 Cumulation of Doses (continued) 1.14E+05 = Units Conversion Factor 10~ pCi 1000 ml 1 yr (2.2-4) 1 pCi 1 liter 8760 hrs 21 = . Adult fish consumption rate (from Table E-5 of Regulatory Guide 1.109, Rev. 1), kg/yr; 730 = Adult water consumption rate (from Table E-5 of Regulatory Guide 1.109, Rev. 1), liters/yr.

D Dilution factor for the drinking water pathway

13. 95 BF, = Bioaccumulation factor for nuclide "1" in fish (from Table A-1 of Regulatory Guide 1.109, Rev. 1), pCi/kg per pci/1 DFL, = Dose conversion factor for nuclide "i" for adults for a particular organ x (from Table E-11 of Regulatory Guide 1. 109, Rev. 1), mrem/pCi Radiological decay constant of \

nuclide "i, " hr ';

0.693 (ta),

(t4) g Radiological half-fife of nuclide "i, " hr; Average transport time to reach point of exposure, hr; 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The more limiting decay time for the drinking water and fish exposure pathways (Reg. Guide 1 ~ 109, Appendix A, Rev. 1).

-ni ~p 2.2-1 presents Table the +, values for an adult receptor. Values of are presented in Table 2.1-4 for each nuclide i.

2.2.2 Com arison A ainst Limits The sum of the cumulative dose from all batch and any continuous releases for a quarter is compared to one-half the design objectives for total body and any organ. The sum of the cumulative doses from all releases for a calendar year is compared to the design objective doses.

The following relationships should hold for the SHNPP to show compliance with ODCM Operational Requirement 3.11.1.2.

2-15

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 2.2.2 Com arison A ainst Limits (continued)

For the calendar quarter:

D~c 1.5 mrem total body (2.2-5) 5 mrem any organ (2.2-6)

For the calendar year:

Dz c 3 mrem total body (2.2-7)

D~ s 10 mrem any organ (2.2-8) where:

D< = Cumulative total dose to any organ t or the total body from all 'releases, mrem:

The quarterly limits given above represent one-half the annual design objective of 10 CFR 50, Appendix I, Section II.A. If any of the limits in equations (2.2-5) through (2.2-8) are exceeded, a special report pursuant to SHNPP Technical Specification 6.9.2 must be filed with the NRC. This report complies with Section IV.A of Appendix I, 10 CFR 50.

The calculations described in Section 2.2.1 will be used to ensure compliance with the limits in 10 CFR 50 Appendix I on a "per-release" basis. The NRC approved program LADTAP II will be used to demonstrate compliance on an annual basis. While there are substantial differences between the ODCM methodology and the methodology in LADTAP, conservative factors have been chosen for the ODCM methodology which will ensure compliance with the limits of 10 CFR 50 Appendix I.

The SHNPP ODCM uses a "modified" NUREG 0133 equation with conservative assumptions. It calculates the dose to a single maximum (ALARA) individual. It does not calculate doses in a steady-state environment, nor does it calculate reconcentration factors. The ALARA individual is an individual that consumes fish caught in the Harris Lake (dilution of 1.0) and receives their drinking water from Lillington, North Carolina (dilution 13.95) .

The SHNPP LADTAP II program uses the guidance of Reg Guide 1.109 and Reg Guide 1. 113 to calculate the dose to a maximum individual, in all age groups, downstream receptors, and integrated populations.

calculate doses based on a steady-state environment using the It will completely-mixed model and reconcentration factors, as per Reg Guide 1.113, and will use average, rather than minimum values for the Cape Fear River flow.

2-16

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 2.2.2 Com arison A ainst Limits (continued)

LADTAP Parameters are maintained in the Laboratory and Facility Services Section Instruction RC-ER-36, Operating instruction for XOQDOQ, GASPAR/

and LADTAP Computer Programs. Parameters subject to change on a periodic basis (flows, source terms, etc.) are not listed, but will be defined and described in the Radioactive Effluent Release Report for the period of interest.

2.2.3 Post-release Com liance With 10 CFR 50 Based ODCM 0 erational Re uirement After the release is made, actual concentrations are used to check 10CFR50 limits, and the actual dilution flow and waste flow are used instead of the anticipated dilution flow and waste flow.

For batch releases, the duration is determined from the actual start and end dates and times of the release. This is used with the actual volume input to calculate the release rate.

2.2.4 Pro'ection of Doses Dose projections for this section are required at least once per 31 days (monthly) in ODCM Operational Requirement 4.11.1.3.1 whenever the liquid radwaste treatment systems are not being fully utilized.

The doses will be calculated using Equation 2.2-1, and projected using the following expression: \

pc ' < p) + ac (2. 3-1) where:

D~, = the 31 Day Projected Dose by organ x Dr sum of all open release points in mrem/day by organ x.

p = the Projection Factor which is the result of 31 divided by the number of days from start of the quarter to the end of the release.

D = Additional Anticipated Dose for liquid releases by organ r and quarter of release.

NOTE: The 31 Day Projected Dose values appear on the Standard and Special Permit Reports. The 31 day dose projections on the Approval/Results screen include any additional dose.

Where possible, expected operational evolutions (i.e., outages, increased power levels, major planned liquid releases, etc.) should be accounted for in the dose projections. This may be accomplished by using the source-term data from similar historical operating experiences where practical, and adding the dose as Additional Anticipated Dose.

2-17

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 2.2.4 Pro ection of Doses (continued)

To show compliance with ODCM Operational Requirement 3.11.1.3, the projected 31 day dose should be compared to the following limits:

D~, s 0.06 mrem for total body (2 ~ 3-2) and D~, s 0.2 mrem for any organ (2. 3-3)

Xf the projections exceed either Expressions 2.3-2 or 2.3-3, then the appropriate portions of the liquid radwaste treatment system shall be used to reduce releases of radioactivity.

2-18

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 LIABLE 'M LIQUID EFFLUENT RELEASE TANKS AND PUMPS

"'o. of Tanks PUMP DESIGN CAPACITY lgpm)

Discharge Recirculation Eductor Factor

"'ank Tank Capacity I al)

Radiation Effluent Monitor ID SWST 100 100 0.2 25,000 REM-3542 WECT 35 35 1.0 10,000 REM-3541 WMTo> 100 100 0.25 25,000 REM-3542 TL&HS 35 100 0.25 25,000 REM-3540 The settling basin has two pumps. When one pump is running, the design flow rate is 500 gpm. When both pumps are running, the design flow rate is 800 gpm.

Reference SHNPP FSAR Tables 11.5.1-1 and 11.2.1-7 SWST: Secondary Waste Sample Tank WECT: Waste Evaporator Condensate Tank WMT: Waste Monitor Tank TLSHS: Treated Laundry and Hot Shower Tank Waste Monitor Tank are used to batch release secondary waste effluent when activity is suspected in this pathway..

2-19

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 XQLE2M Setpoints for Cooling Tower Blowdown Dilution Flow Rates (F,)

Setting Trip Flow Rate (gpm) Minimum Dilution Flow Rate

( pm) 4000 a 6% 3,800 7 000 2 6 6,650 11 000 g 5% 10,450 15 000 g 5% 14,260 Lh.BLE2J -'K Signal Processor Time Constants (T) for GA Technologies RD-53 Liquid Effluent Monitors Detector Back round (cpm) (min) 10 103/cpm bk 10 - 10 103/cpm bk 10 -10 10'/cpm bk 0.01 10'-10'0'-10' 0.01 2-20

Shearon Harris Nuclear Power Plant (SHNPP) December, 1997 Offsite Dose Calculation Manual (ODCM) Rev. 10 TABLE 2.1-4 Page 1 of 3 Nuclide Parameters Half-Life A.t Sensitivity Sensitivity Nuclide (hours) (hr-1) e-Xt (cpm/pCi/ml ) Slope TB Bldg.

Drain Only (cpm/pCi/ml)

H-3 6.46E+06 1.07E-07 1.00E+00 O.OOE+00 O.OOE+00 O.OOE+00 C-14 3.01E+09 2.30E-10 1.00E+00 O.OOE+00 O.OOE+00 O.OOE+00 F-18 1.10E+02 6.32E-03 9.27E-01 O.OOE+00 O.OOE+00 7.78E+07 Na-24 9.00E+02 7.70E-04 9.91E-01 9.36E+07 9.00E-01 9.11E+07 P-32 2.06E+04 3.36E-05 1. OC:.. '00 0. OOE+00 O.OOE+00 O.OOE+00 Cr-51 3.99E+04 1.74E-05 1.00E+00 1.61E+07 1.55E-01 2.79E+06 Mn-54 4.50E+05 1.54E-06 1.00E+00 1.03E+08 9.90E-01 4.45E+07 Mn-56 1.55E+02 4.47E-03 9.48E-01 1.01E+08 9.71E-01 6.41E+07 Fe-55 1.42E+06 4.88E-07 1.00E+00 O.OOE+00 O.OOE+00 O.OOE+00 Fe-59 6.43E+04 1.08E-05 1.00E+00 1.26E+08 1.21E+00 4.58E+07 Co-57 3.90E+05 1.78E-06 1.00E+00 O.OOE+00 O.OOE+00 5.82E+06 Co-58 1.02E+05 6.80E-06 1.00E+00 1.46E+08 1.40E+00 5.68E+07 Co-60 2.77E+06 2.50E-07 1.00E+00 1.89E+08 1.82E+00 9.07E+07 Ni-63 5.27E+07 1.32E-08 1.00E+00 O.OOE+00 O.OOE+00 O.OOE+00 Ni-65 1.51E+02 4.59E-03 9.46E-01 2.24E+07 2.15E-01 1.96E+07 Cu-64 7.62E+02 9.10E-04 9.89E-01 5.16E+07 4.96E-01 1.46E+07 Zn-65 3.52E+05 1.97E-06 1.00E+00 5.24E+07 5.04E-01 2.41E+07 Zn-69 5.56E+01 1.25E-02 8.61E-01 2.22E+03 2.13E-05 5.00E+02 Zn-69m 8.26E+02 8.39E-04 9.90E-01 O.OOE+00 O.OOE+00 3.52E+07 Br-82 2.12E+03 3.27E-04 9.96E-01 O.OOE+00 O.OOE+00 1.43E+08 Br-83 1.43E+02 4.85E-03 9.43E-01 1.95E+06 1.88E-02 5.74E+05 Br-84 3.18E+01 2.18E-02 7.70E-01 6.50E+07 6.25E-01 5.06E+07 Br-85 2.87E+00 2.42E-01 '5.51E-02 6.76E+06 6.50E-02 3.21E+06 Rb-86 2.69E+04 2.58E-05 1.00E+00 8.39E+06 8.07E-02 3.96E+06 Rb-88 1.78E+01 3.89E-02 6.27E-01 1.45E+07 1.39E-01 1.83E+07 Rb-89 1.54E+01 4.50E-02 5. 83E-01 1. 22E+08 1.17E+00 7.00E+07 Sr-89 7.28E+04 9.52E-06 1.00E+00 1.46E+04 1.40E-04 6.72E+03 Sr-90 1.50E+07 4.628-08 1.00E+00 O.OOE+00 O.OOE+00 O.OOE+00 Sr-91 5.70E+02 1.22E-03 9.86E-01 8.16E+07 7.85E-01 3.48E+07 Sr-92 1.63E+02 4.25E-03 9.50E-01 1.01E+08 9.71E-01 4.61E+07 Y-90 3.85E+03 1.80E-04 9.98E-01 O.OOE+00 O.OOE+00 O.OOE+00 Y-91 8.43E+04 8.22E-06 1 . OOE+00 2 . 83E+05 2.72E-03 1.36E+05 Y-91m 4.97E+01 1.39E-02 8.46E-01 1.28E+08 1.23E+00 3.96E+07 Y-92 2.12E+02 3.27E-03 9.62E"01 2.76E+07 2.65E-01 1.17E+07 Y-93 6.06E+02 1.14E-03 9.86E-01 1.37E+07 1.32E-01 3.96E+06 Zr-95 9.22E+04 7.52E-06 1.00E+00 1.07E+08 1.03E+00 4 . 35E+07 Zr-97 1.01E+03 6.86E-04 9.92E-01 2.68E+07 2.58E-01 9.16E+06 Nb-95 5.05E+04 1.37E-05 1.00E+00 1.06E+08 1'.02E+00 4.41E+07 Nb-97 7.21E+01 9.61E-03 8. 91E-01 0 . OOE+00 O.OOE+00 4.33E+07 Mo-99 3.96E+03 1.75E-04 9.98E-01 3.47E+07 3.34E-01 9.38E+06 77 2-21

Shearon Harris Nuclear Power Plant (SHNPP) December, 1997 Offsite Dose Calculation Manual (ODCM) Rev. 10 TABLE 2.1-4 Page 2 of 3 Nuclide Parameters (continued)

Half-Life Xt Sensitivity Sensitivity Nuclide (hours) (hr-1) (cpm/pCi/ml ) Slope TB Bldg.

Drain Only (cpm/pCi/ml)

Tc-101 1.42E+01 4.88E-02 5.57E-01 1.66E+08 1.60E+00 2.92E+07 RU-103 5.67E+04 1.22E-05 1.00E+00 1.38E+08 1.33E+00 3.83E+07 Ru-105 2.66E+02 2.61E-03 9.69E-01 1.71E+08 1.64E+00 5.21E+07 Ru-106 5.30E+05 1.31E-06 1.00E+00 4.S2E+07 4.35E-01 1.43E+07 A -110m 3.60E+05 1.93E-06 1.00E+00 3.22E+08 3.10E+00 1.41E+08 Sn-113 2.76E+03 2.51E-04 9.97E-01 3.08E+06 2.96E-02 4.28E+05 Sb-124 8.67E+04 7.99E-06 1.00E+00 1.59E+08 1.53E+00 8.31E+07 Sb-125 1.46E+06 4.75E-07 1.00E+00 1.21E+08 1.16E+00 3.20E+07 Te-125m 8.35E+04 8.30E-06 1.00E+00 3.00E+05 2.88E"03 1.17E+04 Te-127m 1.57E+05 4.41E-06 1.00E+00 1.33E+04 1.28E-04 6.29E+03 Te-127 5.61E+02 1.24E-03 9.85E-01 1.97E+06 1. 89E-02 4. 14E+05 Te-129m 4.84E+04 1.43E-05 1.00E+00 5.17E+06 4.97E-02 1.95E+06 Te-129 6.96E+01 9.96E-03 8.87E-01 1.58E+07 1.52E-01 4.02E+06 Te-131m 1.80E+03 3.85E-04 9.95E-01 2.17E+08 2.09E+00 7.37E+07 Te-131 2.50E+01 2.77E-02 7.17E-01 OE+08 1.44E+00 2.58E+07 Te-132 4.69E+03 1.48E-04 9.98E-01 1 .39E+08 1.34E+00 1.69E+07 I-130 7.42E+02 9.34E-04 9.89E-01 4.13E+08 3.97E+00 1.41E+08 I-131 1.16E+04 5.98E-05 9.99E-01 1.55E+08 1.49E+00 3.21E+07 1-132 1.38E+02 5.02E-03 9.42E-01 3.31E+08 3.18E+00 1.30E+08 I-133 1.25E+03 5.55E-04 9.93E-01 1.39E+08 1.34E+00 4.28E+07 I-134 5.26E+01 1.32E-02 8.54E-01 3.08E+08 2.96E+00 1.31E+08 I-135 3.97E+02 1.75E-03 9.79E-01 1.03E+08 9.90E-01 5.82E+07 Cs-134 1.08E+06 6.42E-07 1.00E+00 2.60E+08 2.50E+00 9.68E+07 Cs-136 1.90E+04 3.65E-OS 1.00E+00 3.37E+08 3.24E+00 1.11E+08 Cs-137 1.59E+07 4.36E-08 1.00E+00 1.04E+08 1.00E+00 3.90E+07 Cs-138 3.22E+01 . 2.15E-02 7.72E-01 1.15E+08 1.11E+00 8.43E+07 Ba-139 8.31E+01 8.34E-03 9.05E-01 2.34E+07 2.25E-01 2.17E+06 Ba-140 1.84E+04 3.778-05 1.00E+00 6.01E+07 5.78E-01 1.45E+07 Ba-141 1.83E+01 3.79E-02 6.35E-01 2. 53E+08 2.43E+00 S.42E+07 Ba-142 1.07E+01 6.48E-02 4.60E-01 1.47E+08 1.41E+00 4.44E+07 La-140 2.41E+03 2.88E-04 9.97E-01 1.53E+08 1.47E+00 9.06E+07 La-142 9.54E+01 7.27E-03 9.17E-01 9.59E+07 9 '2E-01 7.75E+07 Ce-141 4.68E+04 1.48E-05 1.00E+00 6.11E+07 5.88E-01 4.29E+06 Ce-143 1.98E+03 3.50E-04 9.96E-01 9.60E+07 9.23E-01 1.90E+07 Ce-144 4.09E+05 1.69E-06 1.00E+00 1.30E+07 1.25E-01 7.96E+05 Pr-143 1.95E+04 3.55E-05 1.00E+00 1.08E+02 1.04E-06 5.27E-01 Pr-144 1.73E+01 4.01E-02 6.18E-01 1.68E+06 1.62E-02 1.14E+06 Nd-147 1.58E+04 4.39E-05 9.99E-01 2.86E+07 2.75E-01 8.08E+06 Hf-181 1.02E+03 6.81E-04 9.92E-01 2.08E+08 2.00E+00 4.14E+07 W-187 1.43E+03 4.85E-04 9.94E-01 1.04E+08 1.00E+00 3.09E+07

+ 7 2-22

Shearon Harris Nuclear Power Plant (SHNPP) December, 1997 Offsite Dose Calculation Manual (ODCM) Rev. 10 TABLE 2.1-4 Page 3 of 3 Nuclide Parameters (continued)

Half-Life Xt Sensitivity Sensitivity Nuclide (hours) (hr-1) e-Xt (cpm/pCi/ml) Slope TB Bldg.

Drain Only (cpm/pCi/ml)

Ar-41 1.10E+02 6.30E-03 9. 27E-01 9.28E+07 8.92E-01 4.51E+07 Kr-83m 1.10E+02 6.30E-03 9.27E-01 O.OOE+00 O.OOE+00 O.OOE+00 Kr-85 5.64E+06 1.23E-07 1.00E+00 6.20E+05 5.96E-03 1.75E+05 Kr-85m 2.69E+02 2.58E-03 9.70E-01 1.20E+08 1.15E+00 1.12E+07 Kr-87 7.63E+01 9.08E-03 8.97E-01 9.19E+07 8.84E-01 3.22E+07 Kr-88 1.70E+02 4.08E-03 9.52E-01 7.49E+07 5.19E+07 '.20E-01 Kr-89 3.16E+00 2.19E-01 7.19E-02 1.39E+08 1.34E+00 6.52E+07 Kr-90 5.39E-01 1.29E+00 1.99E-07 1.59E+08 1.53E+00 5.43E+07 Xe-131m 1.70E+04 4.08E-05 1.00E+00 2.62E+06 2.52E-02 2.21E+05 Xe-133 7.55E+03 9.18E-05 9.99E-01 9.90E+04 9.52E-04 9.33E+03 Xe-133m 3.1SE+03 2.20E-04 9.97E-01 1.59E+07 1.53E-01 2.02E+06 Xe-135 5.47E+02 1.27E-03 9.85E-01 1.47E+08 1.41E+00 2.10E+07 Xe-135m 1.54E+01 4.50E-02 5.83E-01 1.14E+08 1.10E+00 3.30E+07 Xe-137 3.83E+00 1.81E-01 1.14E-01 4.85E+07 4.66E-01 1.32E+07 Xe-138 1.41E+01 4.92E-02 5.54E-01 1.20E+08 1.15E+00 4.25E+07 G-ALPHA 1.00E+05 6.93E-06 1.00E+00 O.OOE+00 O.OOE+00 O.OOE+00 G-BETA 1.00E+09 6.93E-10 1.00E+00 O.OOE+00 O.OOE+00 O.OOE+00 Notes to Table 2.1-4 SENSITIVITY 80% of weighted response to 100 - 1400 keV gammas for offline and an adjacent to line monitor which are sodium iodide (NaI) detectors (reference GA Manual E-115-904, June 1980, and Figure 5, Expected Energy Response Normalized for one gamma per disintegration, Drawing 0360-8934 Rev A, page 14, respectively) . Abundances for each gamma from "Radioactive Decay Tables" by David C. Kocher (Report DOE/TIC-11026, Washington, D.C., 1981)

SLOPE The Liquid Effluent Monitor Gamma Sensitivities for nuclide i, relative to Cs-137. To make nuclide i relative to Cs-137, the nuclide sensitivity is divided by the Cs-137 sensitivity. This column does not apply to TB Drains monitor.

2-23

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 Sheet 1 of 3 TABLE 2.2-1 Ag, VALUES FOR THE ADULT FOR THE SHEARON HARRZS NUCLEAR POWER PLANT Af p= 1.1 4E+0 5( + 2 1BF~)DF~,e (mrem/hr per pCi/ml)

Nuclide Bone Liver ~T. Bod ~Th roid ~Kidne Luncu Gr-LLI H-3 O.OOE+00 8.54E-01 8.54E-01 8.54E-01 8.54E-01 8.54E-01 8.54E-01 C-14 . 3.13E+04 6.27E+03 6.27E+03 6.27E+03 6.27E+03 6.27E+03 6.27E+03 NA-24 2.40E+02 2.40E+02 2.40E+02 2.40E+02 2.40E+02 2.40E+02 2.40E+02 P-32 4.52E+07 2.81E+06 1.75E+06 O.OOE+00 O.OOE+00 O.OOE+00 5.08E+06 CR-51 O.OOE+00 O.OOE+00 1 ~ 28E+00 7.63E-01 2.81E-01 1.69E+00 3.21E+02 MN-54 O.OOE+00 4.41E+03 8.41E+02 O.OOE+00 1.31E+03 O.OOE+00 1.35E+04

'. MN-56 O.OOE+00 4.44E+00 7.87E-01 O.OOE+00 5.63E+00 O.OOE+00 1.42E+02 FE-55 6.76E+02 4.67E+02 1.09E+02 O.OOE+00 O.OOE+00 2.60E+02 2.68E+02 FE-59 1.06E+03 2.49E+03 9.54E+02 O.OOE+00 O.OOE+00 6.95E+02 8.29E+03 CO-57 O.OOE+00 2.20E+01 3.66E+01 O.OOE+00 O.OOE+00 O.OOE+00 5.58E+02 CO-58 O.OOE+00 9.33E+01 2.09E+02 O.OOE+do O.OOE+00 O.OOE+00'.OOE+00 1 '9E+03 CO-60 O.OOE+00 2.69E+02 5.94E+02 O.OOE+00 O.OOE+00 5.06E+03 NZ-63 3.20E+04 2.21E+03 1.07E+03 O.OOE+00 O.OOE+00 O.OOE+00 4.62E+02 NZ-65 4.76E+00 6.19E-01 2.82E-01 O.OOE+00 O.OOE+00 O.OOE+00 1.57E+01 CU-64 O.OOE+00 5.45E+00 2.56E+00 O.OOE+00 1.37E+01 O.OOE+00 4.64E+02 ZN-65 2.32E+04 7.39E+04 3.34E+04 O.OOE+00 4.94E+04 O.OOE+00 4.65E+04 ZN-69 6.25E-03 1.20E-02 8.32E-04 O.OOE+00 7.77E-03 O.OOE+00 1.80E-03 ZN-69M 4.46E+02 1.07E+03 9.79E+01 O.OOE+00 6.48E+02 O.OOE+00 6.54E+04 BR-82 O.OOE+00 O.OOE+00 1.81E+03 O.OOE+00 O.OOE+00 O.OOE+00 2.07E+03 BR-83 O.OOE+00 O.OOE+00 1.24E+00 O.OOE+00 O.OOE+00 O.OOE+00 1.79E+00 BR-84 O.OOE+00 O.OOE+00 8.07E-06 O.OOE+00 O.OOE+00 O.OOE+00 6.33E-11 BR-85 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 RB-86 O.OOE+00 9.95E+04 4.63E+04 O.OOE+00 O.OOE+00 O.OOE+00 1.96E+04 RB-88 O.OOE+00 1.94E-10 1.03E-10 O.OOE+00 O.OOE+00 O.OOE+00 2.67E-21 RB-89 O.OOE+00 1.62E-12 1.14E-12 O.OOE+00 O.OOE+00 O.OOE+00 9.43E-26 SR-89 2.38E+04 O.OOE+00 6.84E+02 O.OOE+00 O.OOE+00 0 'OE+00 3.82E+03 SR-90 5.91E+05 O.OOE+00 1.45E+05 O.OOE+00 O.OOE+00 O.OOE+00 1.71E+04 SR-91 1.84E+02 O.OOE+00 7.43E+00 O.OOE+00 O.OOE+00 O.OOE+00 8.77E+02 SR-92 7.84E+00 O.OOE+00 3.39E-01 O.OOE+00 O.OOE400 O.OOE+00 1.55E+02 2-24

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 Sheet 2 of 3 TABLE 2.2-1 (Continued)

A r VALUES FOR THE ADULT FOR THE SHEARON HARRIS NUCLEAR POWER PLANT (mrem/hr per pCi/ml)

Nuclide Bone Liver T~Bod T~hroid ~Kidne Luncu Gr-LLI Y-90 5.57E-01 O.OOE+00 1.49E-02 O.OOE+00 O.OOE+00 O.OOE+00 5.91E+03 Y-91M 2.61E-07 O.OOE+00 1.01E-08 O.OOE+00 O.OOE+00 O.OOE+00 7.67E-07 Y-91 9.24E+00 O.OOE+00 2.47E-01 O.OOE+00 O.OOE+00 O.OOE+00 5.09E+03 Y-92 5.29E-03 OiOOE+00 1.55E-04 O.OOE+00 O.OOE+00 O.OOE+00 9.27E+01 Y-93 7.75E-02 O.OOE+00 2.14E-03 O.OOE+00 O.OOE+00 O.OOE+00 2.46E+03 ZR-95 4.20E-01 1.35E-01 9.12E-02 O.OOE+00 2.11E-01 O.OOE+00 4.27E+02 ZR-97 1.42E-02 2.87E-03 1.31E-03 O.OOE+00 4.34E-03 O.OOE+00 8.90E+02 NB-95 4.43E+02 2.47E+02 1.33E+02 O.OOE+00 2.44E+02 0 'OE+00 1.50E+06 NB-97 3.70E-03 9.36E-04 3.42E-04 O.OOE+00 1.09E-03 O.OOE+00 3.45E+00 MO-99 O.OOE+00 1.14E+02 2.17E+01 O.OOE+00 2.58E+02 O.OOE+00 2.64E+02 TC-99M 2.60E-03 7.35E-03 9.36E-02 O.OOE+00 1.12E-01 3.60E-03 4.35E+00 TC-101 5.81E-18 8.37E-18 8.21E-17 O.OOE+00 1.51E-16 4.28E-18 2.52E-29 RU-103 5.49E+00 O.OOE+00 2.37E+00 O.OOE+bo 2.10E+01 O.OOE+00 6.41E+02 RU-105 7.07E-02 O.OOE+00 2.79E-02 O.OOE+00 9.13E-01 O.OOE+00 4.32E+01 RU-106 ().23E+01 O.OOE+00 1.04E+01 O.OOE+00 1.59E+02 O.OOE+00 5.33E+03 AG-110M 9.55E-01 8.83E-01 5.25E-'01 O.OOE+00 1.74E+00 O.OOE+00 3.60E+02 SB-124 2.33E+01 4.40E-01 9.24E+00 5.65E-02 O.OOE+00 1 '2E+01 6.62E+02 SB-125 1.50E+01 1.67E-01 3.57E+00 1 '2E-02 O.OOE+00 1.16E+01 1.65E+02 TE-125M 2.57E+03 9.32E+02 3.44E+02 7.73E+02 1.05E+04 O.OOE+00 1.03E+04 TE-127M 6.51E+03 2.33E+03 7.94E+02 1.66E+03 2.65E+04 O.OOE+00 2.18E+04 TE-127 4.36E+01 1.57E+01 9.44E+00 3.23E+01 1.78E+02 O.OOE+00 3.44E+03 TE-129M 1.10E+04 4.10E+03 1.74E+03 3.77E+03 4.59E+04, O.OOE+00 5.53E+04 TE-129 2.33E-02 8.76E-03 5.68E-03 1.79E-02 9.80E-02 O.OOE+00 1.76E-02 TE-131M 1.27E+03 6.19E+02 5.16E+02 9.80E+02 6.27E+03 O.OOE+00 6.14E+04 TE-131 4.07E-08 1.70E-08 1.28E-08 3 '5E-08 1.78E-07 O.OOE+00 5.76E-09 TE-132 2.19E+03 1.41E+03 1.33E+03 1.56E+03 1.36E+04 O.OOE+00 6.69E+04 I-130 1.62E+01 4.77E+01 1.88E+01 4.05E+03 7.45E+01 O.OOE+00 4.11E+01 I-131 1.67E+02 2.39E+02 1.37E+02 7.84E+04 4.10E+02 O.OOE+00 6.31E+01 I-132 2.29E-01 6.12E-01 2.14E-01 2.14E+01 9.75E-01 O.OOE+00 1.15E-01 I-133 4.00E+01 6.95E+01 2.12E+01 1.02E+04 1.21E+02 O.OOE+00 6.25E+01 I-134 3.37E-04 9.15E-04 3.27E-04 1.59E-02 1.46E-03 O.OOE+00 7.98E-07 I-135 5.29E+00 1.38E+01 5.11E+00 9.13E+02 2.22E+01 O.OOE+00 1.56E+01 2-25

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 Sheet 3 of 3 TABLE 2.2-1 (Continued)

Air VALUES FOR THE ADULT FOR THE SHEARON HARRES NUCLEAR POWER PLANT (mrem/hr per pCi/ml)

Nuclide Bone Liver ~T. Bod ~Th roid ~Kidne ~nnn Gr-LLI CS-134 2.99E+05 7.10E+05 5.81E+05 0 ~ OOE+00 2.30E+05 7.63E+04 1.24E+04 CS-136 3.05E+04 1.20E+05 8.65E+04 O.OOE+00 6.69E+04 9.17E+03 1.37E+04 CS-137 3.83E+05 5.23E+05 3.43E+05 O.OOE+00 1.78E+05 5.91E+04 1.01E+04 CS-138 4.92E-05 9 '2E-05 4.82E-05 O.OOE+00 7.14E-05 7.06E<<06 4.15E-10 BA-139 3.72E-03 2.65E-06 1.09E-04 O.OOE+00 2.48E-06 1.50E-06 6.60E-03 BA-140 3.08E+02 3.86E-01 2.02E+01 O.OOE+00 1.31E-01 2.21E-01 6.33E+02 BA-141 1.05E-12 7.94E-16 3.55E-14 O.OOE+00 7.38E-16 4.51E-16 4.95E-22 BA-142 1.84E-21 1.89E-24 1.16E-22 0.00E+00 1.60E-24 1.07E-24 O.OOE+00 LA-140 1.34E-01 6.75E-02 1.78E-02 O.OOE+00 O.OOE+00 O.OOE+00 4.96E+03 LA-142 4.51E-05 2.05E-05 5.11E-06 O.OOE+00 O.OOE+00 O.OOE+00 1 '0E-Ol CE-141 7.76E-02 5.24E-02 5.95E-03 O.OOE+00 2.44E-02 O.OOE+00 2.01E+02 CE-143 1.07E-02 7.94E+00 8.79E-04 O.OOE+00 3.50E-03 O.OOE+00 2.97E+02 CE-144 4.08E+00 1.71E+00 2.19E-01 O.OOE+00 1.01E+00 O.OOE+00 1.38E+03 PR-143 5.91E-01 2.37E-01 2.93E-02 O.OOE+00 1.37E-01 O.OOE+00 2.59E+03 PR-144 5.88E-16 2.44E-16 2.99E-17 O.OOE+00 1.38E-16 O.OOE+00 8.46E-23 ND-147 4.02E-01 4.64E-01 2.78E-02 O.OOE+00 2.71E-01 O.OOE+00 2.23E+03 W-187 2.10E+02 1.75E+02 6.12E+01 O.OOE+00 O.OOE+00 O.OOE+00 5.74E+04 NP-239 3.08E-02 3.03E-03 1.67E-03 O.OOE+00 9.44E-03 O.OOE+00 6.21E+02 2-26

0 0 M th 8

{rs Ds I . N rt 0

(({

U X 0

I

{D n Ds Figure 2.M 0 Liquid Weste Processing Flow Dfegrem c n l

Dr {D ken Q. tt Dl Censors{roy Slarele 0 Tents Seeenasry Wo ale 0 Condone Pohhor ere Senora yank (SWS tj Z ((r Ds 8

Dsrelr ersrssl Rates H{prr rt Srseeto dypsos 8 rk tense Wee>> I . ~ S Nonhrlsd O rt tanks a Chere>>ky 4ek

{Wssn 0 C g {A Drabs 4 keen4ry 4 Hol Sooner A De>>speal Drshs Ssssp tanks S ~ Q r{4 tree>>4 M

4eonary 4 I .

Pbor Drsh Shor Dreh Dorsh Hol Skooer tanks leaks Serape rt44847) I I

I Eosdsssn{

Drabs 0

Se>>p Wesle Den>>

Eseporehr Wosle froporelor CenAnsL'o tonk

{WEC77 Resysh CVCS Ds>>h koreans Ho{rap Tanks Son+

Rosyele f parshr re

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 Figure 2.1-2 Liquid Effluent Flow Stream Diagram

~Le end Treated Laundry & Treated Laundry 8 Tank or Basin Hot Shower Tank Hot Shower Tank RBA Radiation Ettiuent Monitor Pump REM REM-1WL4540 Secondary Waste T Sample Tank 0 W

e REM-21WS.3542 r REM B I

0 0 Spcctade Flange W d

Waste Monitor Waste Monitor Waste Neutrallzatlon 0

Tank Tank Basin ~

W n

Waste Evaporator Waste Evaporator Settling Condensate Tank Condensate Tank Basin

~ Liquids containing radioactive matertals other than tritium are not permitted REM by procedure to be sent to the Waste Lake REM-21 WL4541 Neutralization Basin.

2-28

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offside Dose Calculation Manual (ODCM) Rev.

Figure 2.1-3 Normal Service Water Flow Diagram Normal Senrtce Vrater Reactor Auxtttary Building Koat Loddo Koat Loado REM 0400 I A

Vrasta Processing Koat Loado Building REM aaoo 8

trtatn Condenser Turtrtne Buitdlng Normal Service Water Circulating Water Pumps 0 Laoand 0 REM m RadlauOn Etnuent MOnltOr NSW ~ Nonaal Srrvke Water Cooting Tower Basin OO OO NSW Pumps CTMV Pampa CTMV Strainer nyaaaa Ltna Gact$ wash Nants Lake Notdt For detailed trow drawtnsa or tne strake water Syalral, refer lo tna FSAlt 2-29

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 Figure 2.1-4 Other Liquid Effluent Pathways Outside Tank Area Drains Turbine Building Floor Drains ENuent Line ENuent Line Outside Tank Turbine Building Area Drains Floor Drains Sumps REM 3528 REM 3530

~4 Yard Oil Separator Storm Drain System Waste Neutralization ~

Basin Settling Basin Harris Lake Cooling Tower Blowdown Line Turbine Bulfdlng Floor Brains Effluent can be diverted to the Secondary Waste Treatment System

• 'utside Tank Area DraIns Effluent can be diverted to the Llquld Radwaste Treatment Sysyem 2-30

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3.0 GASEOUS EFFLUENTS At SHNPP there are four gaseous effluent discharge points: Plant Vent Stack 1, Turbine Building Vent Stack 3A, and the Waste Processing Building Vent Stacks 5 and 5A. These are shown in Figures 3.1, 3.2, and 3.3 along with their tributaries. All gaseous effluent releases at the plant are considered ground releases.

3.1 Monitor Alarm Set oint Determination (ODCM Operational Requirement 3.3.3.11)

This section provides the methodology for stack effluent monitor setpoints to ensure that the dose rates from noble gases at the site boundary do not exceed the limits of 500 mremlyear to the whole body or 3000 mremlyear to the skin as specified in ODCM Operational Requirement 3.11.2.1.  !

The radioactivity effluent monitors for each stack and for specific effluent streams are shown in Figures 3.1 and 3.3 and are listed in Appendix C.

Gamma spectroscopy analysis of the gas sample should provide the nuclide identification and activity. However, in the case where the noble gas activities are < LLD the relative nuclide composition can be assumed from the GALE code activities foz projected normal operating releases (Table'3.1-1). The GALE code is used to establish a default setpoint for each vent stack. This setpoint will be used as a "fixed" setpoint until a more conservative setpoint is capitulated, either using a different assumed mix or actual sample results.

3-1

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 3.1.1 Default Continuous Release Monitor Set pints Usin a Conservative mix (GALE code)

The following methodology is the default setpoint for the continuous release vent stacks based on conservative assumptions of mix (GALE code) and maximum stack flow rate.

Determine the noble gas. radionuclide activity (Q<) in pCi, .and the activity release rate (Q) in pCi/sec for each nuclide i. Q is the release rate of nuclide v, in pci/sec.

i in gaseous effluent from discharge point Qr, gCq ~ Fduration ~ 28316. 85 (3.1-1a) and Cf Fy 28316 85 / 60 (3.1-1b) where:

index over all vent stacks Cg concentration of nuclide, in pCi/cc the GALE code activities from Table 3.1-1.

effluent release rate or\ vent flow rate in CFM

,the maximum effluent design flow rate at the point of discharge (acfm) from Table 3. 1-3.

duration duration of r'elease, in m'nutes 28316.85 conversion factor for cc/ft~

60 seconds per minute 3-2

Shearon Harris Nuclear Power plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 3.1.1 Default Continuous Release Monitor Set pints Usin a Conservative mix (GALE code) (continued)

Determine the maximum whole body and skin dose rate (mrem/year) during the release.

Q.< = ~(X Q) ( <, K, Q, l (3.1-2a) and Qa-s ~XQ) [ Z~ (L~ w 1.1M~) Q~ ) (3.1-2b) where:

index over all nuclides the total body dose factor due to gamma emissions for Kg noble gas radionuclide Table 3.1-2.

i (in mrem/yr per pCi/m3), from Lq The skin dose factor due to beta emissions for noble gas radionuclide i (mrem/yr per pCi/m~)

Mg air dose factor due to gamma emissions for noble gas radionuclide *i (mrad/yr per pCi/m') . A unit 1M'he conversion constant of 1.1 mrad/mrem converts air dose to skin dose Lg+1. Skin dose factor (mrem/yr per pCi/m'), from Table 3.1-.2

~XQ The highest calculated annual average relative concentration for any sector at or beyond the exclusion boundary (sec/m')

6.1E-06 sec/m~ from Table A-l, Appendix A 3-3

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 3.1.1 Default Continuous Release Monitor Set pints Usin a Conservative mix (GALE code) (continued)

3. Determine the ratio of dose rate limit to dose rate.

nratio = lesser of the ratios Whole Body ratio = 500 (3.1-3a)

Q.a and skin ratio = 3000 (3.1-3b) where:

500 site dose rate limit for whole body in mrem/year.

3000 site dose rate limit for skin in mrem/year.

Determine S~, the maximum concentration setpoint in pCi/cc, and RR the maximum release rate setpoint in pCi/sec for the monitor.

S ( f fn ~ nratio ~

/CD ) + Bkg (3. 1-4a) and RR = S ~

P ~ 28316.852 /,60 (3.1-4b) where safety factor for the discharge point 0.5 fn ~ dose rate allocation factor for the discharge point fraction of the radioactivity from the site that may be released via the monitored pathway to ensure that the site boundary limit is not exceeded by simultaneous releases. These values are based on the percentage of an individual stack flow to the total stack flow and are in Table 3.1-3.

Bkg Monitor background, in pCi/cc 0 for calculation of default setpoint.

3-4

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 3.1.1 Default Continuous Release Monitor Set pints Usin a Conservative mix (GALE code) (continued)

Using the GALE code activities from Table 3.1-1 and the maximum effluent design flow rate, continuous release stack maximum setpoints in pCi/cc and pCi/sec are determined. These values will be used as default values for the stack monitors. Based on sampling and analysis, the setpoint will be recalculated. Zf the sample analysis setpoint is higher than the default setpoint, the setpoint will not be changed. lf the sample analysis setpoint is lower than the default, the setpoint will be changed to reflect the more conservative setpoint. When the setpoint changes again, the more conservative setpoint, comparing the default (GALE code) and sample analysis, will be used.

Determine S<< the gas channel alert alarm setpoint in pCi/cc, and RR, the gas channel alert alarm release rate setpoint in pci/sec.

Salert [ (S~- Bkg) Ar] + Bkg (3. 1-Sa) and alert ( (RR - Bkgrr) Ai) + Bkgrr (3.1-5b) where:

Ar A, value < 1.0 designed to alert the operator that the high alarm setpoint is being approached.

Bkgrr = Bkg P~ 28316. 85 / 60

\

3-5

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 3.1.2 Monitor Set oints Usin Sam le Results In Stacks 1 and 5, the potential exists for batch releases concurrent with the normal continuous ventilation flow of effluents. The sources of batch releases for the Plant Vent Stack 1 include containment normal and pre-entry purge and pressure relief. Batch release sources for Vent Stack 5 include releases from the waste gas decay tanks (WGDT). In these cases, the monitor setpoint must reflect the contribution of both the continuous and batch sources.

The following methodology will calculate a setpoint foz the continuous release vent stacks based on actual sample zesults and for batch releases occurring concurrently with continuous releases.

Determine the noble gas. radionuclide activity (Q<) in pCi, .and the activity release rate (g) in pCi/sec for each nuclide i. Q is the avezage release rate of nuclide i in gaseous effluent from discharge point v, in pCi/sec. Noble gases may be averaged over a period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Qz P Cq ~

F ~ duzation ~ 28316.85 (3.2-1a) and Q>> Cg ~

F ~ 28316.85 / 60 (3.2-1b) where:

index over all vent stacks Cg concentration of nuclide, in pCi/cc the measured concentration from a stack effluent sample or pre-release sample. If there is no activity in the sample, then the GALE code activities from Table 3.1-1 will be used.

effluent release rate or vent flow rate in CPM for continuous releases, the measured effluent flow rate or the maximum effluent design flow rate at the point of release (acfm) from Table 3.1-3.

3-6

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 3.1.2 Monitor Set pints Usin Sam le Results (continued) for batch releases, the release flow rate, in acfm 1,500 acfm for containment normal purge 37, 000 acfm for containment preentry purge

,2. 26 8+06 ( hP 14 ~ 7 t

)

273' T

)

for a containment pressure release 600 ( .'

~ 7

( 2731 T

)

for a Waste Gas Decay Tank release where:

2.26E+06 and 600 are the volumes in containment and decay tank, respectively, and ft'f the T T BP and QP~ are the estimated, respective temperature and change in pressure (psig) following the release of the containment and decay tank; and, 14.7 lb/in', i.e., 1 atmosphere pressure t = Length of release, min 273'K = O'C T~,Ta = 273 K +

duration of release, in minutes C'uration 28316.85 conversion factor for cc/ft~

3-7

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 3.1.2 Monitor Set pints Usin Sam le Results (continued)

2. Determine the maximum whole body and skin dose rate (mrem/year) during the release by summing together the dose rates for this release with all concurrent releases for the time of the release.

Qa-e = ,~(X Q[ Z[ Zq Kq Q~ [ (3. 2-2a) and TXXQ Z[ Zi (Li + 1 ~ LMq[ Q( j (3.2-2b) where:

index over all radionuclides the total body dose factor due to gamma emissions for noble gas radionuclide Table 3.1-2.

i (mrem/yr per pCi/m3), from The skin dose factor due to beta emissions for noble gas radionuclide i (mrem/yr per pCi/m')

The air dose factor due to gamma emissions for noble gas radionuclide i (mrad/yr per pCi/m') . A unit conversion constant of 1.1 mrad/mrem converts air dose to skin dose Lg+l. 1M'kin dose factor (mrem/yr per pCi/m') from Table 3.1-2

~X Q = "

The h'ighest calculated annual average relative concentration for any sector at or beyond the exclusion boundary (sec/mQ) 6.1E-06 sec/m~ from Table A-l, Appendix A 3-8

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 3.1.2 Monitor Set pints Usin Sam le Results (continued)

3. Determine the ratio of dose rate limit to dose rate.

nratio = lesser of the ratios Whole Body ratio = 500 (3.2-3a) and skin ratio = 3000 (3.2-3b) where:

500 site dose rate limit for whole body in mrem/year.

3000 = site dose rate limit for skin in mrem/year.

Determine S~, the maximum concentration setpoint in pCi/cc, and RR the maximum release rate setpoint in pCi/sec for the monitor.

S ( f fn ~ nratio ~

/CD ) + Bkg (3.2-4a) and RR x Smx 'v '8316 852 / 60 (3.2-4b) where safety factor'or the discharge point 0.5 falloc dose rate allocation factor for the discharge point fraction of the radioactivity from the site that may be released via the monitored pathway to ensure that the site boundary limit is not exceeded by simultaneous releases. These values are based on the percentage of an individual stack flow to the total stack flow and are in Table 3.1-3.

Bkg Monitor backgroundI in pci/cc measured background at time of release or 0.

3-9

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 3.1.2 Monitor Set pints Usin Sam le Results (continued)

5. Determine S the gas channel alert alarm setpoint in pCi/cc, and RR, the gas channel alert alarm release rate setpoint in pci/sec.

[ (S~- Bkg) A~) + Bkg (3.2-5a) and RR~g~ [ (RR x Bkgce) Ac] + Bkg (3.2-5b) where:

A value ( 1.0 designed to alert the operator that the high alarm setpoint is being approached.

Bkg Bkg ~

Fy ~ 28316.85 / 60 3-10

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 3.1.3 Effluent Monitorin Durin Ho in O erations If the reactor has been shut down for greater than 30 days, the condenser vacuum pump discharge during initial hogging operations at plant start-up and prior to turbine operation may be routed as dual exhaust to (1) the Turbine Vent Stack 3A and (2) the atmosphere directly. Xn this instance, the blind flange on the latter exhaust route will be removed (see Figure 3.3).

A conservative effluent channel setpoint has been established for or Ven t St ack 3A. The monitor setpoint should be reduced proportionately to the estimated fraction of the main condenser effluent flowing directly to the atmosphere.

3-11

0 (n Table 3.l-l M ra l(r pf GASEOUS SOURCE TERMS rs 0 lir Condenser U X 0

Plant Vent Vacuum Pump WPB WPB Containment (a Ventilation Ventilation Ventilation Ventilation Purge or WGDT (d) fa Flow Flow Flow Flow Pressure Relief Release O (a via via via via via via 1 n c Stack 1 Stack 3A Stack 5" Stack 5A Stack 1 Stack 5 C ()

nr (c) (e)

Ci Ci Ci 0

~lid ~i/c ~rcir ~rc rrf 0

Kr.85m 6.89E-10 7.02 4.69E-9 7.69 2.61E.9 7.02 9.80E-B 2.93 Z pr ra Z

Kr 85 1.87E-7 97.16 6.17E-B 1.85 2.91E-3 97.16 C Kr.87 6.89E.10 7.02 4.69E-9 7.69. 0 2.61E-9 7.02 3.45E-B 1.03 KPBB 1.21E.09 12.28 7.04E.9 11.54 4.56E-9 12.28 1.23E-7 3.69 0 0 n.

U O ~(n Xe.131m 5.17E-10 5.26 2.35E-9 3.85 5.49E-9' 1.96E-9 5.26 3.27E-7 9.77 8.51E-5 2.84 g Xe-133m 0 0 7.26E-B 2.17 x Xe-133 2.76E.09 28.07 1.64E-B 26.92 0 1.04E-B 28.07 1.82E.6 54.29 3.36E-7 4.67E-3 Xe-135m 5.17E-10 5.26 4.69E-9 7.69 1.96E-B 5.26 7.26E 9 0.22 1

Xe-135 .2.93E-9 29.82 1.88E-B 30.77 0 1.11E-B 29.82 7.99E-7 23.89 Xe-138 5.17E-10 5. 26 2.35E.9 3.85 0 1.96E-9 5.26 5.45E-9 0.16 0 (a) Source terms are from SHNPP FSAR Table 11.3.3-1 and not actual releases.

Values apply only to routine releases and not emergency situations.

(b) Source term for this eff(vent stream not presented with FSAR. RAB mix assumed.

(c) Ci (uCi/cc) = Ci r r 5 56 5rnin 1 6 ir i f 0 c (Flow Rate f(3/mm) z0 (d) Based on an assumed flow rate of 15 CFM for WGDT. C (e) Based on an assumed flow rate of 37,000 CFM.

Shearon Harris Nuclear Power Plant (SHNPP) 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE CONSTANTS'ugust 3.1-2 DOSE FACTORS AND Whole Body Dose Skin Dose Factor Factor (Kq) (Lg + l. 1 Mg)

Radionuclide (mrem/yr/pci/m') (mr em/ yr/pCi/m~)

Kr-83m 7.56E"02 2.12E+01 Kr-85m 1.17E+03 2.81E+03 Kr-85 1.61E+01 1 '6E+03 Kr-87 5.92E+03 1.65E+04 Kr-88 1.47E+04 1.91E+04 Kr-89 1.66E+04 2.91E+04 Kr-90 1.56E+04 2.52E+04 Xe-131m 9.15E+01 6.48E+02 Xe-133m 2.51E+02 1.35E+03 Xe-133 2.94E+02 6.94E+02 Xe-135m 3.12E+03 4.41E+03 Xe-135 1.81E+03 3.97E+03

.Xe-137 1.42E+03 1.39E+04 Xe-138 8.83E+03 1.43E+04 Ar-41 8.84E+03 1.29E+04

• Regulatory Guide 1.109, Rev. 1, Table B-l, multiplied by (1.0+E6 pCi/pCi) 3-13

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.1-3 GASEOUS MONITOR PARAMETERS PVS-1 TBVS-3A WPBVS-5 WPBVS-5A Maximum effluent design 390, 000 28, 620 232,500 103,050 flow rate, (acfm)

Flow Allocation Factor 0.517 0.038 0.308 0.137 (f...)

3-14

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3.2 Postrelease Com liance with 10CFR20-Based ODCM 0 erational Re uirement 3.11.2 3.2.1 Noble Gases The gaseous effluent monitors setpoints are utilized to show prerelease compliance with ODCM Operational Requirement 3.11.2.1. However, because they may be based upon a conservative (GALE 'code) mix of radionuclides, when using Table 3.1-1, the possibility exists that the setpoints could be exceeded and yet 10CFR20-based limits may actually be met. Therefore, the following methodology has been provided in the event that high alarm setpoints are exceeded, a determination may be made as to if the whether the actual releases have exceeded the dose rate limits of ODCM Operational Requirement 3.11.2.1.

The dose rate in unrestricted .areas resulting from noble gas effluents is limited to 500 mrem/year to the total body and 3000 mrem/year to the skin. Based upon NVREG-0133, the following equations are used to show compliance:

Ki(~XQ) v Qi (3. 2-1)

~ V s 500 mrem/yr (Li + 1'1 Mi ~XQ)v Qiv s 3000 mrem/yr (3.2-2) i where':

(7X Q)y The highest calculated annual average relative concentration for long-term vent stack releases for areas at or beyond the exclusion boundary sec/m'.

6.1E-06 sec/m'rom Table A-1, Appendix A, for ground-level releases in the S sector at the exclusion boundary.

Ki The total body dose factor due to gamma emissions for noble gas radionuclide "i, " mrem/year per pCi/m'.

Table 3.2-3.

The skin dose factor due to beta emissions for noble gas radionuclide "i, " mrem/year pez pCi/m'.

Table 3.2-3. ~

Mg The air dose factor due to gamma emissions for noble gas radionuclide "i, " mrad/year per pCi/m'.

Table 3.2-3.

l 3-15

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3.2. 1 Noble Gases (continued)

The ratio of the tissue to air absorption coefficients over the energy range of the photon of interest.

Converts mrad to mrem (Reference NUREG-0133). The factors (Lz + 1.1 M~) are tabulated in Table 3.1-2.

iv The release rate of radionuclide "i" in gaseous effluents from all plant vent stacks (pCi/sec) .

The determination of the controlling location for implementation of dose rate limits for noble gas exposure is a function of the historical annual average meteorology.

The radionuclide mix is based on the sampling and analysis required by ODCM Operational Requirement 4.11.2.1.2. Zf the analysis is < LLD, then the GALE code, historical data for the mix, or a Xe-133 / Kr-85 LLD mix for that analysis will be used to demonstrate compliance.

The release rate is derived from either the actual flow rate or the default flow rate and the known or assumed mix.

, Release Rate (pCi/sec) = Flow (cc/sec)

• Concentration (pCi/cc)

The noble gas radionuclide mix was based upon source terms calculated using the NRC GALE Code and presented in the SHNPP FSAR Table 11.3.3-1.

They are reproduced in Table 3.2-1 as a function of release point.'

The X/Q value utilized in the equations is the highest long-term annual average relative concentration (~X Q)v in the unzestricted area for the period 1976 1987. Long-tenn annual average (~X Q) v values at other locations shown in Table 3.2-2 are presented in Appendix A. A description of their derivation is also provided in this appendix.

To select the limiting location for ground-level releases, long-term annual average (~X Q) values were calculated assuming no decay, undepleted transpozt to the exclusion boundary. These values are given in Table A-l, Appendix A ., The maximum exclusion boundary (~XQ) v for ground-level releases occurs in the S sector. Therefore, the limiting location for implement;ation of the dose rate limits for noble gases is considered to be "the exclusion boundary (1.36 miles) in the S sector.

Values for Kq, L~, and M~ which are to be used by SHNPP in Equations 3.2-1 and 3.2-2 to show compliance with ODCM Operational Requirement 3.11.2 are presented in Table 3.2-3. 'These values were taken from Table B-1 of NRC Regulatory Guide 1.109, Revision 1. The values have been multiplied by 1.0E+06 to convert mrad/pCi to mrad/pCi for use in Equations 3.2-1 and 3.2-2.

3-16

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3.2.2 Radioiodines and Particulates The bases for ODCM Operational Requirement 3/4.11.2.1 states that the dose rate to the thyroid of a child in an unrestricted area resulting from the inhalation of radioiodines, tritium, and particulates with half-lives w 8 days is limited to 1500 mrem/yr to any organ. Based upon NUREG-0133, the following is used to show compliance:

P~ [ (~X Q) Q~) c 1500 mrem/yr (3. 2-3) i where:

Pi = The dose paramete" for radionuclides other than noble gases for the inhalation pathway, mrem/yr per pCi/m'.

. In the calculation to show compliance with ODCM Operational Requirement 3.11.2.1.b, only the inhalation pathway is considered.

The radionuclide mix is based on the sampling and analysis required by ODCM Operational Requirement 4.11.2.1.2. If the analysis is < LLD, then no activity is assumed to have been released during the sampling period. The release rate is derived from the flow (actual or default) and the mix.

Release Rate (pCi/sec) = Flow (cc/sec)

• Concentration (pCi/cc)

The determination of the controlling exclusion boundary location was based, upon the highest exclusion boundary (~X Q) value. Values for P

iI in Eq. 3.2-3 were calculated for a child for various radionuclides for the inhalation pathway using S

the methodology of NUREG-0133. The Pi values are presented in Table 3.2-4. A description of the methodology used in calculating the Pz values is presented in Appendix B. The values of P~ reflect, for each radionuclide, the maximum Pz value for any organ.

The (~X Q) v value utilized in Equation 3.2-3 is obtained from the tables presented in Appendix A. A description of the derivation of the X/Q values is provided in Appendix A.

3-17

Table 3.2-1 Q tn Releases from the Shearon Harris Nuclear Power Plant (1) M lO I

Ql Normal Operation (Curies/year)

I w 0 Waste Processing Bldg Exlumst and/or Waste Processing Bldg Exhaust via VENT STACK SA Condenser Vacuum Pump and Turbine Building ax 0

Waste Gas Decay Tanks (2) or RAB/FHB and Contairlment Exhaust lO 8

NOBLE via VENT STACK 5 NORMAL Exhaust via VENT STACK 1 via VENT STACK 3A n

rrn

~S ~SL)'~O~W ~OP <~AEONS C T I XK~B 5XBQ~) I 2',

0 c Kr-85m 0 0 5.4E+01 c n 4.E+00 2.0E+00 6.0E+01 I Kr-85 8.0E+00 6.5E+02 3.4E+01 0 0 6.9E+02 ta Kr-87 0 0 1.9E+01 4.0E+00 2.0E+00 Kr-88 2.5E+01 0 0 0 6.8E+Ol 7.0E+00 3.0E+00 7.8E+01 0

Xe-131m 1.9E+01 1.8E+02 3.0E+00 1.0E+00 2.0E+02 Z IZ Xe-133m 0 4.0E+01 0 0 4.0E+01 Xe-133 0 1.0E+03 1.6E+01 7.0E+00 1.0E+03 c Ql PQ Xe-135m 0 4.0E+00 3.0E+00 2.0E+00 Xe-135 9.0E+00 I I 0 4.4E+02 1.7E+01 8.0E+00 4.7E+02 fU Xe-138 0 3.0E+00 3.0E+00 1.0E+00 7.0E+00 Q a

n-Ar-41 3.4E+00 g V) z'o Adapted from SHNPP FSAR Table 11.3.3-1 and do not reflect actual release data.

These values are only for routine releases and not for a complete inventory of gases in an emergency.

(2) Waste Gas Decay Tank releases assumed after a 904ay decay period.

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 Table 3.2-2 Distance to the Nearest Special Locations for the Shearon Harris Nuclear Power Plant (miles)*

(Comparison of 1993/1994 Data)

Residence Milk Animal Garden Meat Animal Exclusion Sector Boundar 1993 1994 1993 1994 1993 1994 1993 1994 N 1.32 2.2 2.2 2.2 2.2 2.8 2.2 2.2 2.2 NNE 1.33 1.9 1.7 1.9 2.2 2.2 NE 1.33 2.3 2.3 2.3 2.3 2.3 2.3 ENE 1.33 1.8 1.8 3.6 1.8 1.8 1.33 1.8 1.8 2.0 2.0 2.0 ESE 1.33 2.6 2.6 4.7 4.7 SE 1.33 2.6 2.6 2.6 2.6 SSE 1.33 4.2 4.3 1.36 5.3 5.3 5.3 5.3 5.3 SSW 1.33 3.9 3.9 3.9 SW 1.33 2.9 2.9 2.9 2.9 WSW 1.33 4.5 4.5

1. 33 3.0 3.0 3.1 3.1 3.1 3.1 1.33 2.3 2.3 2.3 2.3 2.6 1.26 2.4 . 2.4 2.7 2.7 3.7 3.7

1. 26 1.6 1.6 2.0 2.0 2.0 2.0 As of August, 1994.

Distance estimates are + 0.1 miles except at the exclusion boundary.

Distances and sectors determined by Global Positioning System.

3-19

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.2"3 DOSE FACTORS FOR NOBLE GASES

• Total Body Skin Gamma Air Beta Air Dose Factor Dose Factor Dose Factor Dose Factor Radio nuclide K, L M, N, (mrem/yr per (mrem/yr per (mrad/yr per (mrad/yr per Ci/m') Ci/m~) Ci/m') Ci/m )

Kr-83m 7.56E-02 1.S3E+01 2.88E+ 02 Kr-85m 1.17E+03 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 1.34E+03 1.72E+01 1.95E+03 Kr-87 5.S2E+ 03 9.73E+03 6.17E+03 1.03E+04 Kr-88 1.47E+04 2.37E+03 1.52E+04 2.93E+ 03 Kr-89 1.66E+04 1.01E+04 1.73E+04 1.06E+04 Kr-90 1.56E+04 7.29E+ 03 1.63E+04 7.83E+03 Xe-131m 9.15E+01 4.76E+ 02 1.56E+ 02 1.11E+03 Xe-133m 2.51E+02 9.94E+02 3.27E+02 1.48E+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05E+03 Xe-135m 3.12E+03 7.11E+ 02 3.36E+ 03 7.39E+02 Xe-135 1.81E+03 1.86E+03 1.92E+03 2.46E+03 Xe-137 1.42E+03 1.22E+04 1.51E+03 1.27E+04 Xe-138 8.83E+ 03 4.13E+03 9.21E+03 4.75E+ 03 Ar-41 8.84E+03 2.69E+ 03 9.30E+ 03 3.28E+ 03 The listed dose factors are for radionuclides that may be detected in gaseous effluents.

3-20

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.2-4 PiLI VALUES (INHALATION) FOR A CHILD Pi~I Value ISOTOPE mrem/ r er Ci/m'-3 1.12E+03 P-32 2.60E+06 Cr-51 1.70E+04 Mn-54 1.57E+06 Fe-59 1.27E+06 Co-58 1.10E+06 Co-60 7.06E+06 Zn-'65 9.94E+05 Rb-86 1.98E+05 Sr-89 2.15E+06 Sr-90 1.01E+08 Y-91 2.62E+06 Zr-95 2.23E+06 Nb-95 6.13E+05 Ru-103 6.61E+05 Ru-106 1.43E+07 Ag-110m 5.47E+06 Sn-113 3.40E+05 Sb-124 3.24E+06 Te-127m 1.48E+06 Te-129m 1.76E+06 I-131 1.62E+07 I-132 1.93E+05 I-133 3.84E+06 I-135 7.91E+05 Cs-134 1.01E+06 Cs-136 1.71E+05 Cs-137 9.05E+05 Ba-140 1.74E+06 Ce-141 5.43E+05 Ce-144 1.19E+07 Hf-181 7.95E+05 3-21

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5

3. 3 COMPLIANCE WITH 10CFR50 The calculations described in Section 3.2 will be used to ensure compliance with the limits in 10 CFR 50 Appendix I on a "per-release" basis. The NRC approved program GASPAR will be used to demonstrate compliance on an annual basis. While there are differences between the ODCM methodology and the methodology in GASPAR, conservative factors have been chosen for the ODCM methodology which will ensure compliance with the limits of 10 CFR 50 Appendix I.

The SHNPP ODCM calculates the dose to a single maximum (ALARA) individual. The ALARA individual is an individual that "lives" at the site boundary in the sector that has the most limiting long-term average X/Q value.

The SHNPP GASPAR program calculates the dose to a maximum individual in all age groups, down wind receptors, will calculate and integrated populations. It doses based on the actual average meteorology for the period of interest.

GASPAR Parameters are maintained in the Laboratory and Facility Services Section Instruction RC-ER-36, Operating Instruction for XOQDOQg GASPAR, and LADTAP Computer Programs. Parameters subject to change on a periodic basis (receptor locations, source terms, etc.) are not listed, but will be defined and described in the Radioactive Effluent Release Report for the period of interest.

3-22

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 3.3.1 Noble Gases Cumulation of Doses Based upon NUREG-0133, the air dose in the unrestricted area due to noble gases released in gaseous effluents can be determined by the following equations:

D= 3.17 E-08 P M [ (~XQ) Q v

+ (~Xq) q v]

(3. 3-1)

D = 3 . 17 E-08 p N ( (~X Q) Q v

+ (~X q) qi v') (3. 3-2) where D

Y The air dose from gamma radiation, mrad.

D~ The air dose from beta radiation, mrad.

Ng The air dose factor due to beta emissions for each identified noble gas radionuclide "i," mrad/year per pci/m'.

Table 3.2-3.

~XQ The relative concentration fox areas at or beyond the exclusion boundary for short-term ground-level vent stack releases (c 500 hour0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />s/year), sec/m'. See Section 3.0 earlier or use 6.1E-06 sec/m'rom Table A-l, Appendix A.

Q The total release of noble gas radionuclide "i" in gaseous iv

~

effluents for long term releases (>500 hrs/yr) from all vent stacks (pCi).

q.iv The total release of radionuclide "i" in gaseous releases for short-term releases (F500 hours/year) from all vent stacks, (pCi) .

3.17 E-08 The inverse of the number of seconds in a year (sec/year) '.

Mg The air dose factor due to gamma emissions for each identified noble gas radionuclide (mrad/yr/pCi/m'). A unit conversion constant of 1. 1 mrad/mrem converts air dose to skin dose 3-23

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3.3.1 Noble Gases (continued)

To show compliance with 10CFR50, Expressions'.3-1 and 3.3-2 are evaluated at the controlling location where the air doses are at a maximum.

At SHNPP the limiting location is the exclusion boundary at 1.36 miles

(-2.19 kilometers) in the S sector based upon the tables presented in Appendix A (see Section 3.2.1 earlier). For this document, long-term annual average ~X Qv values can be used in lieu of short-term values (see Section 3.0 earlier).

The determination of the limiting location for implementation of 10CFR50 is a function of parameters such as radionuclide mix,'sotopic release, and meteorology. To select the limiting location, the highest annual average ~X Q value for ground-level releases is controlling. The only source of short-term releases from the plant vent are containment purges, containment pressure relief, and waste gas decay tank release.

Determination of source terms is described in 3.3.1.2.

Values for M~ and Nq, which are utilized in the calculation of the gamma air and beta air doses in Equation 3.3-1 to show compliance with 10CFR50, are presented in Table 3.2-3. These values originate from T able B-1 of the NRC Regulatory Guide 1.109, Revision 1. The values have been multiplied by 1.0E+06 to convert from mrad/pCi to mrad/pCi.

The following relationships should hold for SHNPP to show compliance with ODCM Operational Requirement 3.11.2.2.

For the calendar quarter:

Dy s 5 mrad (3. 3-3)

D5 c 10 mrad (3.3-4)

For the calendar year:

DY s 10 mrad (3.3-5)

D5 s 20 mrad (3.3-6) quarterly limits glen above represent one-half of the aannual si n a d esign jectives of Section II.B.1 of Appendix I of 10CFR50. If any of the ob'he o

limits of Equations 3.3-3 through 3.3-6 are exceeded, a Special Report pursuant to Technical Specification 6.9.2 must be filed with the NRC.

This report complies with Section IV.A of Appendix I of 10CFR50.

3-24

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 3.3.1 Noble Gases (continued)

2. Source Term Determination Containment Batch Purge A purge of containment may be started as a Batch purge and continued as a normal purge. The containment Batch Purge volume is considered to be two air containment volumes (RCB vol = 2.26E+06 ft'). The containment air is sampled and analyzed for noble gases and tritium prior to release.

Stack 1 has a continuous particulate filter and iodine cartridge sampler that is analyzed weekly (minimum) and used for total particulate and iodine effluent accountability for continuous releases. The noble gases and tritium analysis are used for containment effluent accountability as follows; qs = C,+vj (3.3-7)

Where; qq = Activity of nuclide "i" released (pCi).

Cz = Concentration of radionuclide "i" (pCi/cc) v~ = Containment volume (cc) .

Waste Gas Decay Tank Batch Releases, Waste Gas Decay Tanks (WGDT) are sampled and analyzed for tritium and noble gases prior to each release. Stack 5 has a continuous particulate filter and iodine cartridge sampler that is analyzed weekly (minimum) and used for total particulate and iodine effluent accountability for continuous releases. The activity (pCi) for nuclide "i" for Waste Gas Decay Tank effluent accountability is calculated as follows; (Ci ~P~ 600 '8316. 85; 298)

(3.3-7a)

(14. 7 283) 3-25

Shearon Harris Nucleax'ower Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 3.3.1 Noble Gases (continued)

Where; q, Activity of nuclide "i" released (pCi)

Cq = Concentration of nuclide "i" (pCi/cc).

8P, = Change in pressure (psia) of the WGDT (psia = psig + 14.7) 600 = WGDT volume, (ft~)

28316.85 = Conversion factor for converting from ft'o cc.

298 = Sample temperature at time of analysis, (4k) .

14.7 Sample pressure at time of measux'ement, (psia) 283 = WGDT Temperature, 4k (see Note below)

NOTE: The FSAR assumes WGDT temperature to be in the 50-140 4F xange. Since there is no indicator for the actual WGDT temperature, 50'F (10 'C) is conservatively assumed as an acceptable substitute. Also assumed is a sample analysis temperature of 77 F (25 C) .

NOTE: Containment Pressure Releases (ILRT) aze calculated using the same methodology, Containment Pressure Releases are released via Stack 1: The volume to use is 2.26E+06 ft'.

Continuous Releases Each of the four effluent stacks at the HNP have noble gas monitors. Using the net concentration (pCi/cc) from these monitors times the volume released (determined from the flow monitors) the total activity (pCi) of noble gases released are calculated as follows:

Qx Cx 'x (3.3-8)

Where; Q = Total activity (pCi) xeleased fxom Stack "x".

C = Net concentration (pCi/cc) from Stack "x" noble gas monitor.

V= Volume (cc) released from Stack "x" using the flow monitor and, if out of service use the compensatory measurements for volume determination.

3-26

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 n

3.3.1 Noble Gases (continued)

The activity (pCi) released for radionuclide "i" equals the radionuclide "i" fraction of the radionuclide mix times the total activity released from Stack "x".

Qa = Qx '~ (3.3-8a)

Where; C~

s, (3.3-8b) and; S~ = The radionuclide "i" fraction of the radionuclide mix Cz = The concentration of nuclide "i" in the grab sample (pCi/cc).

/CD = Total activity in grab sample (pCi/cc)

The'radionuclide mix is based on the sampling and analysis required by ODCM Operational Requirement 4.11.2.2.1. Zf the grab sample activity is ( LLD, then a mix based on historical data or a mix based on the Xe-133 / Kr-85 LLD mix of that sample may be used.

\

f When a monitor is out of service, the results of the compensatory sampling for each nuclide times the volume released for that time interval will be used for 'effluent accountability. During this situation if the sample shows no detectable activity then there is no activity released.

Corrections for Double Accounting For the two stacks that may have batch releases during the same time interval as continuous releases, the above calculations are corrected for double accounting as follows; Qse=Qi-% (3. 3-9)

Where; Q<, = Total corrected activity of nuclide "i" (pCi) from Stack "x" when batch releases are being made during that time period.

For short term (batch) releases, the effluent stream is sampled and analyzed. The results of the sampling and analysis is used as the source term for the batch release.

Release rate is derived from the source term and the release flow rate.

3-27

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3.3.1 Noble Gases (continued)

3. Projection of Doses Doses resulting from the release of gaseous effluents will be projected once every 31 days (monthly). The doses will be projected utilizing Equations 3.3-1 and 3.3-2, and projected using Dpc where:

')

the following expression:

= (D'K + Dar (3.3-10)

D~, = the 31 Day Projected Dose by organ x Dr sum of all open release points in mrem/day by organ x.

p = the Projection Factor which is the result of 31 divided by the number of days from start of the quarter to the end of the release.

D = Additional Anticipated Dose for liquid releases by organ r and quarter of release.

NOTE: The 31 Day Projected Dose values appear on the Standard and Special Permit Reports. The 31 day dose projections on the Approval/Results screen include any additional dose.

\

Where possible, expected operational evolutions (i.e., outages,

'.increased power levels, major planned batch gas releases, etc. )

should be accounted for ip the dose projections. This" may be accomplished by using the source-term data from similar historical operating experiences where practical, and adding the dose as.

Additional Anticipated Dose.

To show compliance with ODCM Operational Requirement 3.11.2.4, the projected month's dose should be compared as in the following:

Dy ~ 0.2 mrad to air for gamma radiation (3. 3-11) and D5 s 0.4 mrad to air for beta radiation (3.3-12)

Zf the projections exceed either Equations 3.3-11 or 3.3-12, then the appropriate portions of the gaseous radwaste treatment system shall be used to reduce releases of radioactivity.

3-28

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 3.3.2 Radioiodine and Particulates

1. Cumulation of Doses Section ZZ.C of Appendix Z of 10CFR50 limits the release of radioiodines and radioactive material in particulate form from a reactor such that the estimated dose or dose commitment to an individual in an unrestricted area from all pathways of exposure is not in excess of 15 mrem to any organ. Based upon NUREG-0133, the dose to an organ of an individual from radioiodines and particulates with half-lives greater than 8 days in gaseous effluents released to unrestricted areas can be determined by the following equation:

D = 3.17E-08 p (RiJ.Z [(~Xq) v Qi v

+ (X~q) v qi v ]

+

iM iy iG + R B

) [(MDQ) Q + (MDq q ])+

M Z

+ R V

+ R B

~/ Tv 4 (~q) qT v ] ( .3-13) where:

D DoSe to any organ Z from tritium, radioiodines, and particulates, mrem.

The highest long-term (> 500 hr/yr) annual average relative deposition: 8. 8 E-09 m for ~

the food and ground plane pathways at the controlling location which is the exclusion boundary in the S sector (from Table A-4, Appendix A, for ground-level vent stack releases).

(~Dq) v The relative deposition factor for short term, ground-level vent releases (s 500 hrs/yr), m'.

See Section 3.0 earlier if using "real" meteorology or use 8.8 E-09 m from Table A-4, Appendix A, for the food and ground plane pathways at the controlling location.

R. Dose factor for an organ for radionuclide "i" for either the 'cow milk or goat milk pathway, mrem/yr per pci/sec per m ~.

3-29

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 3.3.2 Radioiodine and Particulates (continued)

R ig

= Dose factor for an organ for radionuclide "i" for the ground plane exposure pathway, mrem/yr per pCi/sec per m ~.

Ri = Dose factor for an organ for radionuclide "i" for the inhalation pathway, mrem/yr per pCi/m~.

R.

i= Dose factor for, an or gan for radionuclide "i" for the vegetable pathway, mrem/yr per pci/sec per m'.

ip Dose factor for an organ for radionuclide "i" for the meat pathway, mrem/yr per pCi/sec per m a RT = Dose factor for an organ for tritium for the milk pathway mrem/yr per pCi/m'.

RT = Dose factor for an organ for tritium for the vegetable pathway, mrem/yr per pCi/m'.

RT = Dose factor for an organ for tritium for the inhalation pathway, mrem/yr per pCi/m'.

l Dose factor for an organ for tritium for the meat pathway, mrem/yr per pCi/m'.

Q Tv

='elease o f tritium in gaseous e ffluents for long-term vent stack releases (> 500 hrs/yr),

pCi.

qTv Release of tritium in gaseous effluents for short-term vent stack releases (c 500 hrs/yr),

pCi.

To show compliance with 10CFR50, Equation 3.3-13 is evaluated for a hypothetical individual at the limiting location. At SHNPP this location is the exclusion boundary in the S sector which has the highest annual average ~X Q and MD Q values. This assures that the actual exposure of a member of the public will not be substantially underestimated. The critical receptor is a child.

Appropriate ~X Q and ~D Q values from tables in Appendix A are used. For this document, long-term annual average ~X Q and 7D Q v v values may be used in lieu of short-term values (see Section 3.0 earlier).

3-30

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3.3.2 Radioiodine and Particulates (continued)

The determination of a limiting location for implementation of 10CFR50 for radioiodines and particulates is a function of:

1. Radionuclide mix and isotopic release

2. Meteorology

3. Exposure pathway

4. Receptor's age In the determination of the limiting location, the radionuclide mix of radioiodines and particulates is based on the sampling and analysis required by ODCM Operational Requirement 4.11.2.1.2. If the analysis is < LLD, 'hen no activity is assumed to have been released during the sampling period. The release rate is derived from the flow (actual or default) and the mix.

In the determination of the limiting sector, all age groups and all of the exposure pathways are initially evaluated using the GASPAR code. These include cow milk, beef and vegetable ingestion, inhalation, and ground plane exposure. Goat milk is not currently an exposure pathway at SHNPP.

SHNPP ODCM Operational Requirement 3.12.2 requires that a land-use census survey be conducted on an annual basis. The age groupings at the various receptor locations ake also determined during this survey. Thus, depending on the results of the survey, a new

'limiting location and receptor age group could result.

To avoid possible annual revisions to the ODCM software which evaluates effluent releases for compliance with 10CFR50, the limiting sector location has been fixed at the exclusion boundary in the S sector where the highest historical annual average ~X Qv and ~D Q values occur (Appendix A). With all of the exposure pathways identified in the land use census (Table 3.2-2) available to" a hypothetical receptor, the critical organ is a child's bone.

This approach avoids a substantial underestimate of the dose to a real member of the public.

Long-term MX Q and ~D Q values for ground-level releases are provided in tables in Appendix A. They may be utilized additional special location arises different from those presented if an in the special locations of Table 3.2-2. A description of the derivation of the various X/Q and D/Q values is presented in Appendix A.

3-31

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3.3.2 Radioiodine and Particulates (continued)

Tables 3.3-1 through 3.3-19 present Rz values for the total body, GI-tract, bone, liver, kidney, thyroid, and lung organs for the ground plane, inhalation, cow milk, goat milk, vegetable, and meat ingestion pathways for the infant, child, teen, and adult age groups as appropriate to the pathways. These values were calculated using the methodology described in NUREG-0133 assuming a grazing period of eight months. A description of the methodology is presented in Appendix B.

The following relationship should hold for SHNPP to show compliance with SHNPP ODCM Operational Requirement 3.11.2.3.

For the calendar quarter:

Dx s 7.5 mrem (3.3-14)

For the calendar year:

Dx s 15 mrem (3.3-15)

The quarterly limits given above represent one-half the annual design objectives of Section II.C of Appendix I of 10CFR50. If any of the limits of Equations 3.3-14 or 3.3-15 are exceeded, a Special Report pursuant to Technical Specification 6.9.2 must be filed with the NRC. This report complies,with Section IV.A of Appendix I of 10CFR50.

Projection of Doses Doses 'resulting from release of radioiodines and particulates will be projected once every 31 days (monthly). The doses will be projected utilizing Equation 3.3-13, and projected using, the Dpr = (D'K where:

')

following expression:

+ Der (3. 3-16)

Dpr the 31 Day Projected Dose by organ x Dr sum of all open release points in mrem/day by organ x.

the Projection Factor which is the result of 31 divided by the number of days from start of the quarter to the end of the release.

D.r Additional Anticipated Dose for gaseous releases by organ r and quarter of release.

3-32

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3.3.2 Radioiodine and Particulates (continued)

NOTE: The 31 Day Projected Dose values appear on the Standard and Special Permit Reports. The 31 day dose projections on the Approval/Results screen include any additional dose.

Where possible, expected operational evolutions (i.e., outages, increased power levels, major planned batch gas releases, etc.)

should be accounted for in the dose projections. This may be accomplished by using the source-term data from similar historical operating experiences where practical, and adding the dose as Additional Anticipated Dose.

To show compliance with ODCM Operational Requirement 3.11.2.4, the projected month's dose should be compared as in the following:

D s 0.3 mrem to any organ (3.3-17)

Zf the projections exceed Expression 3.3-14, then the appropriate portions of the gaseous radwaste treatment system shall be used to reduce releases of radioactivity.

3-33

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.3.1 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'GE PATHWAY ~ Ground GROUP ~ ALL bturJida Cr-51 4.66E+06 4.66E+06 4.66E+06 4.66E+06 4.66E+06 4.66E+06 4.66E+06 5.51E+06 Mn-54 1.34E+09 1.34E+09 1.34E+09 1.34E+09 1.34E+09 1.34E+09 1.34E+09 1.57E+09 Fe-59 2.75E+08 2.75E+08 2.75E+08 2.75E+08 2.75E+08 2.75E+08 2.75E+08 3.23E+08 Co-58 3.79E+08 3.79E+08 3.79E+08 3.79E+08 3.79E+08 3.79E+08 3.79E+08 4.44E+09 Co.60 2.15E+ 10 2.15E+ 10 2.15E+ 10 2.15E+ 10 2.15E+ 10 2.15E+ 10 2.15E+10 2.52E+ 10 Zn-65 7.49E+08 7.49E+08 7.49E+08 7.49E+08 7A9E+08 7.49E+08 7.49E+08 8.61E+08 Rb-86 8.99E+06 8.99E+06 8.99E+06 8.99E+06 8.99E+06 8.99E+06 8.99E+06 1.03E+07 Sr-89 2.23E+04 2.23E+04 2.23E+04 2.23E+04 2.23E+04 2.23E+04 2.23E+04 2.58E+04 Y-91 1.08E+06 1.08E+06 1.08E+06 1.08E+06 1.08E+06 1.08E+06 1.08E+06 1.22E+06 Zr-95 2.49E+08 2.49E+08 2.49E+08 2.49E+08 2.49E+08 2.49E+08 2.49E+08 2.89E+08 Nb-95 1.36E+08 1.36E+08 1.36E+08 1.36E+08 1.36E+08 1.36E+08 1.36E+08 1.60E+08 RU-103 1.09E+08 1.09E+08 1.09E+08 1.09E+08 1.09E+08 1.09E+08 1.09E+08 1.27E+08 Ru-106 4.19E+08 4.19E+08 4.19E+08 4.19E+08 4.19E+08 4.19E+ 08 4.1 9E+ 08 5.03E+ 08 Ag-110M 3.48E+09 3.48E+09 3.48E+09 3.48E+09 3.48E+09 3.48E+09 3.48E+09 4.06E+09 Sn-113 1.A4E+07 6.28E+06 1.22E+07 6.21E+06 1.00E+07 1.33E+07 8.14E+06 4.09E+07 Sb-124 8.76E+08 7.53E+08 8.99E+08 .7.76E+08 8.17E+08 1.01E+09 8.23E+08 1.24E+09 Te-127M 9.15E+04 9.15E+04 9.15E+04 9.15E+04 .9.15E+04 9.15E+04 9.15E+04 1.08E+05 Te-129M 2.00E+07 2.00E+07 2.00E+07 2.00E+07 2.00E+07 2.00E+07 2.00E+07 2.34E+07 I-131 1.72E+07 1.72E+07 1.72E+07 1.72E+07 1.72E+07 1.72E+07 1.72E+07 2.09E+07 1-132 1.24E+06 1.24E+06 1.24E+06 1.24E+06 1.24E+06 1.24E+06 1.24E+06 1.46E+06 1-133 2A7E+06 2.47E+06 2.47E+06 2.47E+06 2.47E+06 2.47E+06 2.47E+06 3.00E+06 I-135 2.56E+ 06 2.56E+ 06 2.56E+ 06 2.56E+ 06 2.56E+ 06 2.56E+ 06 2.56E+ 06 2.99E+ 06 Cs-134 6.82E+ 09 6.82E+09 "

6.82E+ 09 6.82E+ 09 6.82E+ 09 6.82E+ 09 6.82E+ 09 7.96E+ 09 Cs-1 36 1.49E+08 1.49E+08 1.49E+08 1.49E+08 1.49E+08 1.49E+08 1.49E+08 1.69E+08 Cs-1 37 1.03E+10 1.03E+10 1.03E+10 1.03E+10 1.03E+10 1.03E+10 1.03E+10 1.20E+10 Ba-140 2.05E + 07 2.05 E+ 07 2.05E + 07 2.05E+ 07 2.05E + 07 2.05E+ 07 2.05E + 07 2.34E+ 07 Ce-141 1.36E+07 1.36E+07 1.36E+07 1.36E+07, 1.36E+07 1.36E+07 1.36E+07 1.53E+07 C0-144 6.95E+ 07 6.95E+ 07 6.95E+ 07 6.95E+ 07 6.95E+ 07 6.95E+ 07 6.95E+ 07 8.03E+ 07 Hf-181 1.97E+ 08 1.63E+ 08 2.30E+ 08 1.70E+ 08 1.77E+ 08 2.33E+ 08 1.82E+08 2.82E+08

'R Values in units of mrem/yr per yCi/m'or inhalation and tritium and in units of mrem/yr per yCi/sec per m't for all others.

3-34

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.3-2 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'GE PATHWAY ~ Veget GROUP ~ Adult XJhu4t JQdttay IhyxairL H-3 2.28E+ 03 2.28E+03 O.OOE+01 2.28E+ 03 2.28E+ 03 2.28E+ 03 2.28E+ 03 2.28E+ 03 P-32 5.91E+ 07 1.72E+ 08 1.53E+09 9.51E+ 07 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Cr-51 4.60E+04 1.16E+07 O.OOE+01 O.OOE+01 1.01E+04 2.75E+04 6.10E+04 O.OOE+01 Mn.54 5.83E+ 07 9.36E+ 08 O.OOE+ Ol 3.05E+ 08 9.09E+07 O.OOE+ Ol O.OOE+ Ol O.OOE+ 01 FQ-59 1.1 2E+ 08 9.75E+ 08 1.24E+ 08 2.93E+ 08 O.OOE+ 01 O.OOE+ 01 8.17E+ 07 O.OOE+ 01 Co-58 6.71E+07 6.07E+08 O.OOE+01 2.99E+07 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Co-60 3.67E+08 3.12E+09 O.OOE+01 1.66E+08 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+ 01 1

Zn-65 5.77E+08 8.04E+08 4.01E+08 1.28E+09 8.54E+08 O.OOE+01 O.OOE+01 O.OOE+ 01 Rb.86 1.03E+ 08 4.36E+ 07 O.OOE+ 01 2.21E+ 08 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Sr-89 2.87E+ 08 1.60E + 09 1.00E+ 1 0 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ Ol O.OOE+01 Sr-90 1.64E+ 11 1.93E+ 10 6.70E+ 11 O.OOE+01 O.OOE+ 01 O.OOE+ 01 O.OOE+ Ol O.OOE+01 Y-91 1.34E+05 2.76E+09 5.01E+06 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Zr-95 2.51E+05 1.17E+09 1.16E+06 3.71E+05 5.82E+05 O.OOE+01 O.OOE+01 O.OOE+01 Nb-95 4.19E+04 4.73E+08 1.40E+05 7.79E+04 7.70E+04 O.OOE+01 O.OOE+01 O.OOE+01 RU-103 2.04E+06 5.53E+08 4.74E+06 O.OOE+01 1.81E+07 O.OOE+01 O.OOE+01 O.OOE+01 Ru-106 2.46E+07 1.26E+10 1.94E+08 O.OOE+01 3.7$ E+08 O.OOE+01 O.OOE+01 O.OOE+01 Ag-110M 6.23E+06 4.28E+09 1.13E+07 1.05E+ 07 2.06E+ 07 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Sn-113 1:36E+07 2.52E+08 1A4E+07 5.66E+05 4.09E+05 1.96E+05 O.OOE+01 O.OOE+01 Sb-124 4.02E+ 07 2.88E+ 09 1.01E+ 08 1:.92E+ 06 O.OOE+ 00 2.46E+ 05 7.90E+ 07 O.OOE+ Ol Te-127M 6.1 2E+ 07 1.68E+ 09 5.02E+ 08 1.80E+08 '.04E+ 09 1.28E+ 08 O.OOE+01 O.OOE+ 01 Te-129M 4.71E+07 1.50E+09 2.98E+08 1.11E+08 1.24E+09 1.02E+08 O.OOE+01 O.OOE+01 I-131 6.61E+ 07 3.04E+ 07 8.07E+ 07 1.1 5E+ 08 1.98E+ 08 3.78E+ 1 0 O.OOE+ 01 O.OOE+ 01 I-132 5.21E+01 2.80E+01 5.57E+Ol 1.49E+02 2.37E+02 5.21E+03 O.OOE+01 O.OOE+01 1-1 33 1.12E+ 06 3.30E+ 06 2.1 1E+ 06 3.67E+06 6.40E+ 06 5.39E+ 08 O.OOE+ 01 O.OOE+ Ol 1-135 3.91E+04 1.20E+05 4.05E+04 1.06E+05 1.70E+05 7.00E+06 O.OOE+01 O.OOE+Ol Cs-134 8.83E+ 09 1.89E+ 08 4.54E+ 09 1.08E+ 10 3.49E+ 09 O.OOE+01 1.16E+ 09 O.OOE+ 01 Cs-136 1.19E+ 08 1.88E+ 07 4.19E+ 07 1.66E+ 08 9.21E+ 07 O.OOE+Ol 1.26E+ 07 O.OOE+ 01 Cs-137 5.94E+ 09 1.76E+08 6.63E+ 09 9.07E+ 09 3.08E+ 09 O.OOE+ 01 1.02E+ 09 O.OOE+ 01 Ba-140 BAOE+06 2.64E+ 08 1.28E+08 1.61E+05 5.47E+04 O.OOE+01 9.22E+04 O.OOE+01 Ce-141 1.48E+04 4.99E+08 1.93E+05 1.31E+05 6.07E+04 O.OOE+01 O.OOE+ Ol O.OOE+ 01 Ce-144 1.69E+ 06 1.06E+ 10 3.15E+ 07 1.32E+ 07 7.80E+ 06 O.OOE+ Ol O.OOE+ Ol O.OOE+ Ol Hf-1 81 1.07E+ 06 7.06E+ 08 9.51E+ 06 5.36E+ 04 4.48E + 04 3.41E+ 04 O.OOE + 01 O.OOE+ 01 I

'R Values in units of mrem/yr per yCI/m~ for inhalation and tritium and in units of mrem/yr per yCi/sec per m'or all others.

3-35

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.3-3 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'GE PATHWAY ~ Veget GROUP ~ Teen LBadg QLIract Ihyxoid H-3 2.61E+03 2.61E+03 O.OOE+ 01 2.61E+ 03 2.61E+ 03 2.61E+ 03 2.61E+ 03 2.61E+ 03 P-32 6.80E+ 07 1.47E+ 08 1.75E+ 09 1.09E+ 08 O.OOE+ 01 O.OOE+ Ol O.OOE+ 01 O.OOE+ 01 Cr-51 6.11E+04 1.03E+07 O.OOE+01 O.OOE+01 1.34E+04 3.39E+04 8.72E+04 O.OOE+01 Mn.54 8.79E+ 07 9.09E+ 08 O.OOE+ 01 4.43E+ 08 1.32E+08 O.OOE+01 O.OOE+01 O.OOE+01 Fe.59 1.60E+08 9.78E+08 1.77E+08 4.14E+08 O.OOE+Ol O.OOE+01 1.30E+08 O.OOE+01 Co.58 9.79E+07 5.85E+08 O.OOE+01 4.25E+07 O.OOE+ 01 O.OOE+ 01 O.OOE+01 O.OOE+ 01 Co-60 5.57E+08 3.22E+09 O.OOE+01 2.47E+08 O.OOE+ 01 O.OOE+01 O.OOE+01 O.OOE+ 01 Zn.65 8.68E+08 7.88E+08 5.36E+08 1.86E+09 1.19E+09 O.OOE+01 O.OOE+01 O.OOE+ 01 Rb-86 1.30E+08 4.09E+07 O.OOE+01 2.76E+08 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ Ol Sr-89 4.36E+08 1.81E+09 1.52E+10 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Sr-90 2.05E+ 1 1 2.33E+ 10 8.32E+ 11 O.OOE+ Ol O.OOE+ 01 O.OOE+ Ol O.OOE+ 01 O.OOE+ Ol Y-91 2.06E+05 3.15E+09 7.68E+06 O.OOE+ Ol O.OOE+ 01 O.OOE+01 O.OOE+01 O.OOE+01 Zr-95 3.68E+05 1.23E+09 1.69E+06 5.35E+ 05 7.86E+ 05 O.OOE+01 O.OOE+01 O.OOE+01 Nb.95 5.77E+04 4A8E+08 1.89E+05 1.05E+05 1.02E+05 O.OOE+01 O.OOE+01 O.OOE+Ol Ru-103 2.90E+06 5.66E+08 6.78E+06 O.OOE+01 2.39E+07 O.OOE+01 O.OOE+01 O.OOE+01 Ru-106 3.93E+07 1.50E+ 10 3.12E+08 O.OOE+ 01 6.07E+08 O.OOE+01 O.OOE+Ol O.OOE p Ol Ag-110M 9.39E+06 4.34E+09 1.63E+07 1.54E+07 2.g5E+07 O.OOE+01 O.OOE+01 O.OOE+ 01 Sn-113 2.02E+07 2.29E+08 1.91E+07 8.03E+05 5.65E+05 2.63E+05 O.OOE+01 O.OOE+01 Sb-1 24 5.89E+07 3.04E+09 1.51E+08 2:78E+06 O.OOE+01 3.43E+05 1.32E+08 O.OOE+01 Te-127M 9.44E+07 1.98E+09 7.93E+08 2.81E+08 3.22E+09 1.89E+08 O.OOE+01 O.OOE+01 Te-129M 6.79E+07 1.61E+09 4.29E+ 08 1.59E+ 08 1.79E+ 08 1.38E+ 08 O.OOE+ 01 O.OOE+ 01 I-1 31 5.77E+07 2.13E+07 7.68E+07 1.07E+08 1.85E+08 3.14E+ 10 O.OOE+01 O.OOE+01 1-132 4.72E+01 5.72E+01 5.02E+01 1.31E+02 2.07E+02 4.43E+03 O.OOE+01 O.OOE+01 I-133 1.01E+06 2.51E+06 1.96E+06 3.32E+06 5.83E+06 4.64E+08 O.OOE+01 O.OOE+01 I-1 35 3.49E+04 1.04E+05 3.66E+04 9.42E+04 1 49E+05 6.06E+06 O.OOE+01 O.OOE+01 Cs-134 7.54E+ 09 2.02E+ 08 6.90E+ 09 1.62E+ 10 5.1 6E+ 09 O.OOE+ 01 1.97E+ 09 O.OOE+ 01 Cs-136 1.13E+08 1.35E+07 .4.28E+07 1.68E+08 9.16E+07 O.OOE+01 1.44E+07 O.OOE+01 Cs-137 4.90E+ 09 2.00E+ 08 1.06E+ 10 1.41E+ 10 4.78E+ 09 O.OOE+ 01 1.86E+ 09 O.OOE+ 01 Ba-140 8.88E+ 06 2.12E+ 08 1.38E+ 08 1.69E+ 05 5.72E+ 04 O.OOE+ 01 1.14E+ 05 O.OOE+ 01 Ce.141 2.12E+ 04 5.29E+ 08 2.77E+ 05 1.85E+ 05 8.70E+ 04 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Ce-144 2.71E+ 06 1.27E+ 10 5.04E+ 07 2.09E+ 07 1.25E+ 07 O.OOE+ 01 O.OOE+ 01 O.OOE+ Ol Hf-181 1.54E + 06 6.90E + 08 1.38E + 07 7.58E + 04 6.32E + 04 4.63E + 04 O.OOE + Ol O.OOE+ 01

'R Values in units of mrem/yr per yCi/ms for inhalation and tritium and in units of mrem/yr per yCi/sec per m't for all others.

3-36

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.3-4 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'GE PATHWAY ~ Veget GROUP ~ Child ody Mraot ~m ~ytoid ~tag H-3 4.04E+ 03 4.04E+ 03 O.OOE+ 01 4.04E+03 4.04E+03 4.04E+03 4.04E+ 03 4.04E+03 P-32 1.42E+08 1.01E+08 3.67E+ 09 1.72E+ 08 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Cr-51 1.16E+05 6.15E+06 O.OOE+ Ol O.OOE+01 1.76E+04 6.44E+04 1.18E+05 O.OOE+01 Mn-54 1.73E+08 5.44E+08 O.OOE+01 6.49E+08 1.82E+08 O.OOE+Ol O.OOE+01 O.OOE+Ol Fe-59 3.17E+08 6.62E+08 3.93E+08 6.36E+08 O.OOE+01 O.OOE+01 1.84E+08 O.OOE+01 Co.58 1.92E+ 08 3.66E+ 08 O.OOE+01 6.27E+ 07 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Co-60 1.11E+09 2.08E+09 O.OOE+ 01 3.76E+08 O.OOE+ 01 O.OOE+01 O.OOE+01 O.OOE+01 Zn.65 . 1.70E+09 4.81E+08 1.03E+09 2.74E+ 09 1.73E+ 09 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Rb.86 2.81E+08 2.94E+07 O.OOE+01 4.56E+ 08 O.OOE+ 01 O.OOE+01 O.OOE+01 O.OOE+01 Sr-89 1.03E+09 1.40E+09 3.62E+10 O.OOE+ 01 O.OOE+Ol O.OOE+01 O.OOE+01 O.OOE+01 Sr-90 3.49E+ 1 1 1.86E+ 10 1.38E+ 12 O.OOE+01 O.OOE+ 01 O.OOE+ 01 O.OOE+ Ol O.OOE+ 01 Y-91 4.89E+ 05 2.44E+09 1.83E+07 O.OOE+Ol O.OOE+01 O.OOE+Ol O.OOE+Ol O.OOE+01 Zr-95 7.44E+05 8.71E+ 08 3.80E+06 8.35E+05 1.20E+06 O.OOE+01 O.OOE+01 O.OOE+ 01 Nb-95 1.12E+05 2.91E+08 4.04E+ 05 1.57E+05 1.48E+05 O.OOE+01 O.OOE+01 O.OOE+ 01 RU-103 5.86E+ 06 3.94E+08 1.52E+07 O.OOE+01 3.84E+07 O.OOE+01 O.OOE+01 O.OOE+01 Ru-106 9.38E+ 07 1.17E+10 7.52E+ 08 O.OOE+ 01 1.0QE+ 09 O.OOE >01 O.OOE+ 01 O.OOE+ 01 Ag-110M 1.87E+07 2.78E+09 3A6E+07 2.34E+ 07 4.35E+07 O.OOE+ 01 O.OOE+ 01 O.OOE+01 Sn-113 3;97E+ 07 1.45E+ 08 3.64E+ 07 1.18E+06 8.09E+ 05 4.82E+ 05 O.OOE + Ol O.OOE+ Ol Sb-124 1.21E+ 08 2.16E+09 3.44E+ 08 4.47E+ 06 O.OOE+ 01 7.61 E+ 05 1.91 E+ 08 O.OOE + 01 Te-1 27M 2.26E+ 08 1.54E+09 1.9OE+09 5.12E+08 5.42E+ 09 4.55E+08 O.OOE+01 O.OOE+01 Te-129M 1.55E+08 1.22E+09 9.98E+ 08 2.79E+ 08 2.93E+09 3.22E+08 O.OOE+01 O.OOE+01 1-1 31 8.16E+ 07 1.23E+ 07 1.43E+ 08 1.44E+ 08 2.36E+ 08 4.75E+ 10 O.OOE+01 O.OOE+01 I-132 7.53E+ 01 1.93E+02 8.91E+01 1.64E+02 2.51E+02 7.60E+03 O.OOE+01 O.OOE+01 1-133 1.67E+ 06 1.78E+06 3.57E 0 06 4.42E+06 7.36E+06 8.21E+08 O.OOE+01 O.OOE+01 I-135 5.54E+04 8.92E+ 04 6.50E+ 04 1.17E+ 05 1.79E+ 05 1.04E+ 07 O.OOE+ Ol O.OOE+ 01 Cs-134 5.40E+ 09 1.38E+08 1.56E+10 2.56E+10 7.93E+09 O.OOE+01 2.84E+09 O.OOE+01 Cs-136 1.43E+08 7.77E+ 06 8.04E+ 07 2.21E+ 08 1.18E+ 08 O.OOE+ 01 1.76E+ 07 O.OOE+ 01 Cs-137 3.52E+ 09 1.50E+ 08 2.49E+ 10 2.39E+ 10 7.78E+ 09 O.OOE+ 01 2.80E+09 O.OOE+01 BQ-140 1.61E+07 1.40E+08 2.76E+ 08 2.42E+ 05 7.87E+ 04 O.OOE+ 01 1.44E+05 O.OOE+01 Ce-141 4.75E+ 04 3.99E+ 08 6.42E+ 05 3.20E+ 05 1.40E+05 O.OOE+01 O.OOE+01 O.OOE+Ol Ce-144 6.49E+ 06 9.94E+ 09 1.22E+ 08 3.81E+ 07 2.11E+07 O.OOE+01 O.OOE+01 O.OOE+Ol Hf-181 3.15E+06 5.17E+08 3.13E+07 1.22E+05 9.78E+ 04 1.03E+ 05 O.OOE+ 01 O.OOE+ 01

'R VaIues in units of mrem/yr per yCi/m'or inhaIation and tritium and in units of mrem/yr per yCi/sec per m'or all others.

3-37

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.3-5 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'ATHWAY

~ AGE GROUP ~ Adult

~ Meat bluciida QlZxaat JQdmy. Ihyraid H-3 3.27E+ 02 3.27E+ 02 O.OOE+01 3.27E+02 3.27E+02 3.27E+ 02 3.27E+ 02 3.27E+ 02 P-32 1.18E+08 3.43E+08 3.05E+09 1.89E+08 O.OOE+01 O.OOE+Ol O.OOE+Ol O.OOE+01 Cr-51 4.27E+ 03 1.08E+ 06 O.OOE+ 01 O.OOE+ 01 9A2E+ 02 2.56E+ 03 5.67E+ 03 O.OOE+ 01 Mn-54 1.06E+06 1.71E+07 O.OOE+ 01 5.57E+06 1.66E+06 O.OOE+01 O.OOE+01 O.OOE+01 Fe.59 1.43E+08 1.25E+09 1.59E+08 3.74E+08 O.OOE+01 O.OOE+Ol 1.04E+08 O.OOE+01 Co-58 2.43E+07 2.20E+08 O.OOE+01 1.08E+07 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Co.60 1.03E+08 8.76E+08 O.OOE+01 4.66E+07 O.OOE+01 O.OOE+Ol O.OOE+01 O.OOE+01 Zn-65 3.58E+08 4.98E+08 2.49E+08 7.91E+08 5.29E+08 O.OOE+01 O.OOE+01 O.OOE+Ol Rb-86"'.42E+08 6.00E+07 O.OOE+01 3.04E+08 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Sr-89 5.23E+ 06 2.92E+ 07 1.82E+ 08 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+01 Sr-90 2.02E+ 09 2.38E+ 08 8.22E+ 09 O.OOE+ 0 1 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Y-91 1.80E+04 3.71E+08 6.75E+05 O.OOE+01 O.OOE+Ol O.OOE+01 O.OOE+Ol O.OOE+01 Zr-95 2.43E+05 1.14E+09 1.12E+06 3.59E+05 5.64E+05 O.OOE+01 O.OOE+01 O.OOE+01 Nb-95 4.12E+05 4.65E+ 09 1,38E+ 06 7.66E+ 05 7.58E+ 05 O.OOE+ 01 O,OOE+ 01 O,OOE+ 01 Ru-103 2.72E+07 7.38E+09 6.32E+07 O.OOE+01 2.41E+08 O.OOE+01 O.OOE+01 O.OOE+01 Ru-106 2.19E+08 1.12E+ 11 1.73E+09 O.OOE+Ol 3.35E+ 09 O.OOE+ 01 O.OOE + 01 O.OOE+ 01 Ag-110M 2.34E+ 06 1.61E+ 09 4.27E+ 06 3.95E+06 7.76E+06 O.OOE+Ol O.OOE+01 O.OOE+01 Sn-113 2.80E+07 5.19E+08 2.97E+07 1.15E+06 8.40E+05 4.03E+05 O.OOE+01 O.OOE+01 Sb-124 4.72E+06 3.38E+08 1.19E+07 2.25E+05 O.OOE+01 2.88E+04 9.27E+06 O.OOE+01 Te-127M 1.00E+08 2.76E+09 8.22E+08 2.94E+08 3.34E+09 2.10E+08 O.OOE+ 01 O.OOE+ Ol Te-129M 1.17E+08 3.73E+09 7,40E+08 2.76E+08 3.09E+09 2.54E+08 O.OOE+ 01 O.OOE+01 1-131 5.77E+ 06 2.66E+ 06 7.04E+06 1.01E+07 1.73E+07 3.30E+09 O.OOE+01 O.OOE+01 1-133 1.51E.01 4.46E-01 2.85E.01 4.96E.01 8.66E-01 7.29E+ 01 O.OOE+ 01 O.OOE+ 01 1-135 6.07E-17 1.86E-16 6.28E-17 1.64E-16 2.64E.16 1.08E-14 O.OOE+01 O.OOE+01 Cs-134 7.81E+08 1.67E+07 4.01E+08 9.55E+08 3.09E+08 O.OOE+01 1.03E+08 O.OOE+01 Cs-136 2.14E+07 3.33E+06 7.53E+06 2.97E+07 1.65E+07 O.OOE+Ol 2.27E+06 O.OOE+01 Cs.137 4.99E+08 1.47E+07 5.57E+08 7.61E+08 2.58E+08 O.OOE+Ol 8.59E+07 O.OOE+01 Ba.140 1.20E+ 06 3.77E+ 07 1.83E+07 2.30E+04 7.82E+03 O.OOE+01 1.32E+04 O.OOE+01 Ce-141 6.46E+02 2.18E+07 8.42E+ 03 5.69E+ 03 2.65E+ 03 O.OOE+01 O.OOE+ 01 O.OOE+ 01 Ce.144 4.70E+04 2.96E+08 8.75E+05 3.66E+05 2.17E+05 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Hf-181 1.52E+06 9.97E+08 1.34E+07 7.57E+ 04 6.33E+ 04 4.81E+ 04 O.OOE+ 01 O.OOE+ 01

'R Values in units of mrem/yr per yci/m for inhalation and tritium and in units of mrem/yr'per yci/sec per m'or all others.

3-38

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.3.6 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'GE PATHWAY ~ Meat GROUP ~ Teen Lady. QLItact JQdneg Ihyrrud H-3 1.95E+02 1.95E+02 O.OOE+01 1.9SE+02 1.95E+02 1.95E+ 02 1.95E+02 1.95E+02 P-32 9.98E+07 2.16E+08 2.58E+09 1.60E+08 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+Ol Cr-51 3.42E+03 5.75E+05 O.OOE+01 O.OOE+Ol 7.49E+02 1.9OE+03 4.88E+03 O.OOE+01 Mn-54 8.43E+05 8.72E+06 O.OOE+01 4.25E+06 1.27E+06 O.OOE+01 O.OOE+Ol O.OOE+01 Fe-59 1.15E+08 7.02E+08 1.27E+08 2.97E+08 O.OOE+01 O.OOE+01 9.36E+07 O.OOE+01 Co-58 1.93E+07 1.15E+08 O.OOE+01 8.36E+06 O.OOE+Ol O.OOE+01 O.OOE+01 O.OOE+01 Co.60 8.15E+ 07 4.71E+ 08 O.OOE+ 01 3.6 "E+ 07 O.OOE+ 01 O.OOE+01 O.OOE+ 01 O.OOE+ Ol Zn-65 2.83E+08 2.57E+08 1.75E+08 6.07E+08 3.89E+ 08 O.OOE+01 O.OOE+01 O.OOE+ 01 Rb-86 1.19E+08 3.76E+07 O.OOE+Ol 2.54E+08 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ Ol Sr-89 4.40E+06 1.83E+07 1.54E+08 O.OOE+01 O.OOE+01 O.OOE+Ol O.OOE+01 O.OOE+01 Sr-90 1.31E+ 09 1.49E+ 08 5.32E+ 09 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Y-91 1.52E+04 2.33E+08 5.68E+05 O.OOE+Ol O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+Ol Zr-95 1.95E+05 6.53E+08 8.97E+05 2.83E+05 4.16E+05 O.OOE+01 O.OOE+01 O.OOE+ 01 Nb-95 3.29E+05 2.55E+09 1.08E+06 5.97E+05 5.79E+05 O.OOE+01 O.OOE+01 O.OOE+01 Ru-103 2.20E+ 07 4.30E+ 09 5.15E+ 07 O,OOE+ Ol 1.82E+ 08 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Ru-106 1.84E+08 7.00E+10 1.46E+09 O.OOE+ 01 2.81 (+ 09 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Ag-110M 1.86E+06 -

8.59E+08 3.23E+ 06 3.06E+ 06 5.85E+06 O.OOE+01 '.OOE+01 O.OOE+01 Sn-113 2.22E+07 2.51E+08 2.09E+ 07, 8.80E+05 6.19E+05 2.88E+05 O.OOE+ 01 O.OOE+ 01 Sb-124 3.80E+06 1.96E+08 9.73E+06 1. jf9E+05 O.OOE+01 2.21E+04 8.50E+ 06 O.OOE+ Ol Te-127M 8.25E+07 1.73E+09 6.94E+08 2.46E+08 2.81E+09 1.65E+08 O.OOE+01 O.OOE+01 Te-129M 9.81E+07 2.33E+ 09 6.20E+08 2.30E+08 2.59E+09 2.00E+08 O.OOE+01 O.OOE+01 I-131 4.40E+06 1.62E+06 5.85E+ 06 8.20E+ 06 1.41E+ 07 2.39E+ 09 O.OOE p 01 O.OOE+ 01 I-133 1.23E-01 3.06E.01 2.39E.01 4.05E-01 7.10E-01 5.65E+01 O.OOE+01 O.OOE+ 01 I-135 4.88E-17 1.46E-16 5.11E-17 1.32E-16 2.08E-16 8.46E-15 O.OOE+01 O.OOE+01 Cs-134 3.48E+08 9.34E+06 3.19E+08 7.51E+08 2.39E+08 O.OOE +01 9.11E+07 O.OOE+01 Cs-136 1.55E+07 1.86E+06 5.87E+06 2.31E+07 1.26E+07 O.OOE+ 01 1.98E+06 O.OOE+Ol Cs-137 2.14E+08 8.75E+06 4.62E+08 6.15E+08 2.09E+08 O.OOE+01 8.13E+07 O.OOE+01 Ba-140 9.76E+05 2.34E+07 1.51E+07 1.86E+04 6.29E+ 03 O.OOE+ 01 1.25E+ 04 0.00E+ Ol Ce-141 5.42E+ 02 1.35E+ 07 7.07E+ 03 4.72E+ 03 2.22E+03 O.OOE+01 O.OOE+01, O.OOE+01 Ce-144 3.96E+04 1.85E+08 7.37E+05 3.05E+05 1.82E+05 O.OOE+ 01 O.OOE+01 O.OOE+ 01 Hf-181 1.22E+06 5.50E+08 1.10E+07 6.05E+04 5.04E+ 04 3.69E+ 04 O.OOE+ 01 O.OOE+ 01 R Vaiues in units of mrcm/yr per /rCI/m~ for inhalation and tritium and in units of mrem/yr per pCi/sec per m'or all others.

3-39

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.3-7 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'ATHWAY

~ Meat AGE GROUP ~ Child H-3 2.36E+02 2.36E+02 O.OOE+01 2.36E+02 2.36E+02 2.36E+02 2.36E+02 2.36E+02 P-32 1.87E+08 1.34E+08 4.86E+09 2.27E+08 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+Ol Cr-51 5.33E+03 2.83E+05 O.OOE+01 O.OOE+01 8.09E+02 2.96E+03 5.40E+03 O.OOE+01 Mn.54 1.30E+06 4.08E+06 O.OOE+01 4.86E+06 1.36E+06 O.OOE+01 O.OOE+01 O.OOE+01 Fe-59 1.82E+08 3.80E+08 2.25E+08 3.65E+08 O.OOE+01 O.OOE+Ol 1.06E+08 O.OOE+01 Co-58 2.99E+ 07 5.70E+ 07 O.OOE F 01 9.76E+ 06 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Co.60 1.27E+ 08 2.38E+ 08 O.OOE+ 01 4.30E+ 07 O.OOE+ 01 O.OOE+ Ol O.OOE+ 01 O.OOE+ 01 Zn-65 4.3SE+08 1.23E+08 2.62E+08 6.99E+08 4.40E+08 O.OOE+01 O.OOE+01 O.OOE+01 Rb-86 2.21E+08 2.32E+07 O.OOE+01 3.60E+08 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Sr-89 8.31E+06 1.13E+07 2.91E+08 O.OOE+Ol O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+Ol Sr-90 1.74E+ 09 9.26E+ 07 6.87E+ 09 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 0 1 Y-91 2.87E+04 1.43E+ 08 1.07E+ 06 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE 4 01 O.OOE+ 01 Zr-95 3.12E+05 3.65E+08 1.59E+06 3.50E+05 5.01E+05 O.OOE+01 O.OOE+01 O.OOE+01 Nb-95 5.17E+05 1.34E+09 1.86E+06 7.23E+05 6.80E+05. O.OOE+01 O.OOE+01 O.OOE+01 Ru-103 3t58E+07 2.41E+09 9.31E+07 O.OOE+01 2.34E+08 O.OOE+01 O.OOE+01 O.OOE+01 Ru-106 3.43E+08 4.27E+10 2.75E+09 O.OOE+01 3.71E+09 O.OOE+01 O.OOE+01 O.OOE+01 Ag-110M 2.89E+06 4.30E+08 5.36E+ 06 3.62E+ 06 6.74E+ 06 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Sn-'113 3.42E+07 1.25E+ 08 3.14E+07 1.01E+06 6.97E+05 4.15E+05 O.OOE+01 O.OOE+01 Sb-124 6.17E+06 1.10E+08 1.76E+07 2.28E4 05 O.OOE+01 3.88E+04 9.77E+06 O.OOE+01 Te-127M 1.55E+08 1.06E+09 1.31E+09 3.52E+08 3.73E+09 3.13E+08 O.OOE+01 O.OOE+01 Te-129M 1.81E+08. 1.42E+09 1.17E+09 3.26E+08 3.43E+09 3.77E+08 O.OOE+Ol O.OOE+01 I-131 6.20E+06 9.72E+05 1.09E+07 1.09E+07 1.79E+07 3.61E+09 O.OOE+Ol O.OOE+01 1-133 2.07E.01 2.21E-01 4.43E-01 5.48E-01 9.13E-01 1.02E+02 O.OOE+01 O.OOE+01 1-135 7.87E-1 7 1.27E-1 6 9.25E-1 7 1.66E-1 6 2.55E-1 6 1.47E-1 4 O.OOE+ 01 O.OOE+ 01 Cs-1 34 1.95E+08 4.93E+06 5.63E+08 9.23E+08 2.86E+08 O.OOE+01 1.03E+08 O.OOE+01 Cs-1 36 1.80E+ 07 9.78E+ 05 1.01E+ 07 2.78E+ 07 1.48E+07 O.OOE+ 01 2.21E+ 06 O.OOE+ Ol Cs-137 1.20E+ 08 5.10E+ 06 8.51E+ 08 8.1 5E+ 08 2.65E+ 08 O.OOE+ Ol 9.55E+ 07 O.OOE+ Ol Ba.140 1.63E+ 06 1.42E+ 07 2.80E+ 07 2.45E+ 04 7.97E+ 03 O.OOE+ 01 1.46E+ 04 O.OOE+ 01 Ce-141 9.86E+ 02 8.28E+ 06 1.33E+ 04 6.64E+03 2.91E+ 03 O.OOE+ 01 O.OOE+ 01, O.OOE+ 01 Ce-144 7.42E+04 1.14E+08 1.39E+06 4.36E+05 2.41E+ 05 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Hf-181 2.02E+06 3.31E+08 2.00E+07 7.79E+ 04 6.26E+04 6.5 6E + 04 O.OOE + 01 O.OOE + 01

'R Values inunits of mrem/yr per yCi/m'or inhalation and tritium and in units of mrern/yr per yCi/seo per m'or all others.

3-40

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.3-8 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'GE PATHWAY $$ Cow Milk GROUP $$ AdUIt H-3 7.69E+02 7.69E+02 O.OOE+01 7.69E+02 7.69E+02 7.69E+02 7.69E+02 7.69E+02 P-32 4.32E+08 1.26E+09 1.12E+10 6.95E+08 O,OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Cr-51 1.73E+ 04 4.36E+06 O.OOE+ 01 O.OOE+ 01 3.82E+ 03 1.04E+ 04 2.30E+ 04 O.OOE+ 01 Mn-54 9.76E+ 05 1.57E+ 07 O.OOE+ 01 5.11 E+ 06 1.52E+ 06 O.OOE+ 01 O.OOE.01 O.OOE+ 01 Fe-59 1.60E+07 1.39E+08 1.77E+07 4.17E+07 O.OOE+Ol O.OOE+Ol 1.17E+07 O.OOE+Ol Co.58 6.28E+06 5.68E+07 O.OOE+01 2.80E+06 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Co-60 2.24E+ 07 1.91E+ 08 O.OOE+ 01 1.02E+ 07 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Zn-65 1.38E+09 1.92E+09 9.59E+08 3.0SE+09 2.04E+09 O.OOE+01 O.OOE+01 O.OOE+01 Rb-86 7.54E+08 3.19E+08 O.OOE+01 1.62E+09 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Sr-89 2.50E+07 1.40E+08 8.70E+08 O.OOE+01 O.OOE+Ol Q.OOE+Ol O.OOE+01 O.OOE+01 Sr-90 7.59E+09 8.94E+08 3.09E+10 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+Ol O.OOE+01 Y-91 1.37E+02 2.81E+06 5.11E+03 O.OOE+01 O.OOE+ 01 O.OOE+ Ol O.OOE+ 01 O.OOE+ 01 t Zr-95 Nb-95 Ru-103 Ru-106 Ag-110M Sn-113 1.22E+02 1.48E+ 04 2.63E+02 1.60E+03 2.04E+ 07 1.32E+06 5.71E+05 1.67E+ 08 7.14E+04 8.17E+05 1.40E+ 10 2,44E+07 5.62E+02 4.95E+04 6.11E+02 1.26E+04 3.71E+ 07 1.40E+06 1.80E+02 2.75E+04 O.OQE+01 O.OOE+01 3A4E+ 07 5.41E+04 2.83E+ 02 2.72E+04 2.33E+03 2.44E+ 04 6.76E+ 07 3.95E+04 O.OOE+ 01 O.OOE+01 O.OOE+01 O.OOE+ 01 O.OOE+ 01 1.90E+04 O.OOE+ 01 O.OOE+01 O.OOE+01 O.OOE+ 01 O.OOE+ 01 O.OOE+01 O.OOE+ 01 O.OOE+01 O.OOE+01 O.OOE+ 01 O.OOE+ 0 1 O.OOE+01 Sb-124 6.14E+06 4.39E+08 1.55E+07 2.92E+05 O.OOE+01 3.75E+04 1.20E+07 O.OOE+01 Te-1 27M 4.1 1 E+ 06 1.13E+ 08 3.37E+ 07 1.21E+ 07 1.37E+ 08 8.62E+ 06 O.OOE+ 01 O.OOE+ 01 Te-129M 6.19E+06 1.97E+08 3.91E+07 1.46E+07 1.63E+08, 1.34E+07 O.OOE+01 O.OOE+01 I-l 31 1.59E+08 7.32E+07 1.94E+08 2.77E+08 4.76E+08 9.09E+10 O.OOE+Ol O.OOE+01 I-132 1.03E+01 5.51E-02 1.10E+01 2.93E+01 4.67E+01 1.03E+01 O.OOE+01 O.OOE+01 I-133 1.40E+06 4.1 3E+ 06 2.64E+ 06 4.59E+ 06 8.01E+ 06 6.75E+ 08 O.OOE+ 01 O.OOE+ 01 I-1 35 9.03E+03 2.76E+04 9.34E+03 2.45E+04 3.92E+04 1.61E+06 O.OOE+01 O.OOE+Ol Cs-1 34 6.71E+ 09 1.44E+ 08 3.45E+ 09 3.21E + 09 2.66E + 09 O.OOE + 01 8.82E + 08 O.OOE+ 01 Cs-136 4.73E+ 08 7.46E+ 07 1.66E+08 6.S7E+ 08 3.65E+ 08 O.OOE+ 01 5.01E+ 07 O.OOE p 01 Cs-137 4.22E+09 1.25E+08 4.71E+09 6.44E+09 2.19E+09 O.OOE+Ol 7.27E+08 O.OOE+01 Ba-140 1.12E+06 3.53E+07 1.71E+ 07 2.15E+ 04 7.32E+ 03 O.OOE+ 01 1.23E+ 04 O.OOE+ Ol Ce-141 2.23E+02 7.52E+06 2.91E+03 1.97E+03 9.14E+02 O.OOE+01 O,OOE+Ol O.OOE+Ol Ce-144 1.15E+ 04 7.26E+ 07 2.15E+ 05 8.97E+ 04 5.32E+ 04 O.OOE+ 01 O.OOE+ 01 O.OOF + 01 Ht-181 6.68E+02 4.38E+05 03 3.33E + 01 2.79E + 01 2.1 2E+ 01 O.OOE+ 01 O.OOE+ 01 pCi/m'.91E+

'R Values in units of mrem/yr per

~

for inhalation and tritium and in units of mrom/yr per pCi/sec per m'or all others.

3-41

Shearon Harris Nuclear Power Plant (SHNPP) ,August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.3-9 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'GE PATHWAY ~ Cow Milk GROUP ~ Teen ZJ3nc4r. JQdm4r H-3 1.OOE+03 1.0OE+ 03 O.OOE+ 01 1.00E+ 03 1.00E+03 1.0OE y03 1.0OE+ 03 1.00E+ 03 P-32 8.0OE+08 1.73E+09 2.06E+10 1.28E+09 O.OOE+01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Cr-51 3.02E+04 5.08E+06 O.OOE+01 O.OOE+01 6.63E+03 1.68E+04 4.32E+04 O.OOE+01 Mn-54 1.69E+06 1.75E+07 O.OOE+01 8.52E+06 2.54E+06 O.OOE+01 O.OOE+01 O.OOE+01 Fe-59 2.79E+ 07 1.71E+08 3.10E+07 7.23E+07 O.OOE+01 O.OOE+Ol 2.28E+07 O.OOE+01 Co.58 1.09E+07 6.50E+07 O.OOE+01 4.72E+06 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Co.60 3.88E+07 2.25E+08 O.OOE+01 1.72E+07 O.OOE+Ol O.OOE+01 O.OOE+01 O.OOE+Ol Zn-65 2.38E+09 2.16E+09 1.47E+09 5.11E+09 3.27E+09 O.OOE+01 O.OOE+01 O.OOE+01 Rb-86 1.39E+09 4.37E+08 O.OOE+01 2.95E+09 O.OOE+01 O.OOE+Ol O.OOE+01 O.OOE+01 Sr-89 4.59E+07 1.91E+08 1.60E+09 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Sr-90 1.08E+ 10 1.23E+ 09 4.37E+ 10 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+01 O.OOE+01 Y-91 2.52E+02 3.85E+06 9.40E+03 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+Ol O.OOE+Ol Zr-95 2.13E+02 7.16E+05 9.83E+02 3.10E+02 4.56E+02 O.OOE+01 O.OOE+Ol O.OOE+Ol Nb-95 2.58E+04 2.00E+08 8.45E+04 4.68E+04 4.54E+04 O.OOE+01 O.OOE+01 O.OOE+Ol Ru-103 4.65E+02 9.08E+04 1.09E+03 O.OOE+ 01 3.83E+ 03 O.OOE+01 O.OOE+Ol O.OOE+01 Ru-106 2.93E+03 1.11E+06 2.32E+04 O.OOE+ 01 4'.48E+04 O.OOE+01 O.OOE+01 O.OOE+Ol Ag-110M 3'.53E+07 1.63E+10 6.14E+07 5.81E+07 1.11E+08 O.OOE+Ol O.OOE+01 O.OOE+01 Sn-113 2.28E+06 2.58E+07 2.15E+06 9.06E+ 04 6.37E+04 2.97E+04 O.OOE+Ol O.OOE+01 Sb-1 24 1.08E+07 5.56E+ 08 2.76E p 07 5.08E+05 O.OOE+ 01 6.26E+04 2.41E+ 07 O.OOE+01 Te-127M 7.39E+06 1.55E+08 6.22E+ 07 2.21E+07 2.52E+ 08 1.48E+07 O.OOE+01 O.OOE+01 Te-129M 1.13E+07 2.69E+08 7.15E+07 2.65E+ 07 2.99E+08 2.31E+ 07 O.OOE+ 01 O.OOE+Ol I-131 2.65E+08 9.75E+ 07 3.52E+ 08 4.93E+ 08 8.48E+ 08 1.44E+ 1 1 O.OOE+Ol O.OOE+ 01 I-132 1.83E+01 2.22E+01 1.94E+01 5.09E+01 8.02E+01 1.71E+01 O.OOE+01 O.OOE+01 I ~ 133 2.49E+06 6.19E+06 4.82E+06 8.18E+06 1.43E+07 1.14E+09 O.OOE+01 O.OOE+01 1-135 1.58E+ 04 4.74E+ 04 1.66E+04 4.27E+04 6.75E+ 04 2.75E+ 06 O.OOE+ 01 O.OOE+ 01 Cs-1 34 6.54E+ 09 1.75E+08 5.99E+ 09 1.41E+ 10 4.48E+09 O.OOE+ 01 1.71E+ 09 O.OOE+ 01 Cs-136 7.48E+ 08 8.97E+ 07 2.83E+ 08 1.1 1E+09 6.07E+ 08 O.OOE+ 01 9.56E+ 07 O.OOE+ 01 Cs-1 37 3.96E+ 09 1.62E+ 08 8.54E+ 09 1.14E+ 10 3.87E+ 09 O.OOE+ 01 1.50E+ 09 O.OOE+ 01 Ba-140 1.99E+06 4.77E+07 3.09E+07 3.79E+04 1.28E+04 O.OOE+01 2.55E+04 O.OOE+01 Ce-141 4.09E+02 1.02E+07 5.33E+03 3.56E+03 1.68E+03 O.OOE+01 O.OOE+01 O.OOE+01 Co-144 2.12E+04 9.93E+07 3.95E+OS 1.63E+05 9.76E+ 04 O.OOE + Ol O.OOE + 01 O.OOE+ 01 Hf-181 1.18E+ 03 S.28E+ 05 1.06E+ 04 5.82E+ Ol 4.84E+ 01 3.55E+01 O.OOE+Ol O.OOE+01

'R Values in units of mrern/yr por /rCI/m~ for inhalation and tritium and in units of mrern/yr per pCi/soc per m'or all others.

3-42

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.3-10 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'ATHWAY

~ Cow Milk AGE GROUP ~ Child LBo4 Xidttay 1.58E+ 03 1.58E+ 03 O.OOE+01 1.58E+03 1.58E+03 1.58E+03 1.58E+03 1.58E+03 P.32 1.96E+09 1.41E+09 5.09E+10 2.38E+09 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Cr-51 6.17E+04 3.27E+06 O.OOE+01 O.OOE+01 9.36E+03 3.42E+04 6.25E+04 O.OOE+01 Mn.54 3.39E+06 1.07E+07 O.OOE+01 1.27E+07 3.57E+06 O.OOE+01 O.OOE+01 O.OOE+01 Fe.59 5.79E+07 1.21E+ 08, 7.18E+ 07 1.16E+08 O.OOE+ Ol O.OOE+ 01 3.37E+ 07 O.OOE+ Ol Co-58 2.21E+07 4.20E+07 O,OOE+01 7.21E+06 O.OOE+01 O.OOE+01 O.OOE+Ol O,OOE+01 Co.60 7.90E+ 07 1.48E+ 08 O.OOE+ 01 2.68E+07 O.OOE+ 01 O.OOE+01 O.OOE+ 01 '.OOE+ 01 Zn.65 4.79E+09 1.35E+09 2.89E+09 7.70E+09 4.85E+ 09 O.OOE+Ol O.OOE+ 01 O.OOE+ 01 Rb-86 3.36E+09 3.52E+08 O.OOE+01 5A7E+09 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+Ol Sr-89 1.13E+08 1.54E+08 3.97E+09 O.OOE+Ol O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Sr-90 1.87E+ 10 9.95E+08 7.38E+ 10 O.OOE+Ol O.OOE+Ol O.OOE+Ol O.OOE+01 O.OOE+01 Y-91 6.21E+02 3.09E+06 2.32E+04 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Zr-95 4.47E+ 02 5.23E+05 2.28E+ 03 5.02E+02 7.1 8E+ 02 O.OOE+Ol O.OOE+ 01 O.OOE+ Ol l

Nb-95 5.31E+04 1.37E+08 1.91E+05 7.42E+04 6.98E+04 O.OOE+01 O.OOE+01 O.OOE+01 Ru-103 9.88E+02 6.65E+04 2.57E+03 O.OOE+Ol 6.47E+

\

03 O.OOE+ 01 O.OOE+ Ol O.OOE+ 01 Ru-106 7.14E+03 8.90E+05 5.72E+ 04 O.OOE+ 01 7.72E+04 O.OOE+01 O.OOE+01 O.OOE+01 Ag-110M 7.19E+07 1.07E+10 1.33E+08 9.00E+07 1.68E+08 O.OOE+01 O.OOE+01 O.OOE+01 Sn-113 4.61E+05 1.69E+06 4.22E+05 146E+ 04 9.37E+ 03 5.58E+ 03 O.OOE+ 01 O.OOE+ Ol Sb.124 2.29E+07 4.09E+08 6.53E+07 8.47E+05 O.OOE+01 1A4E+05 3.62E+07 O.OOE+01 Te-127M 1.82E+07 1.24E+08 1.53E+08 4.13E+07 4.37E+08 3.66E+07 O.OOE+01 O.OOE+ 01 Te-129M 2.74E+07 2.15E+08 1.76E+08 4.92E+07 5.1 8E+ 08 5.68E+ 07 O.OOE+ Ol O.OOE+ 01 I-1 31 4.88E+08 7.64E+07 8.54E+08 8.59E+08 1.41E+09 2.84E+11 O.OOE+01 O.OOE+Ol I-132 3.89E+01 9.95E+01 4.60E+01 8.45E+01 1.29E+00 3.92E+01 O.OOE+Ol O.OOE+01 1-133 5.48E+06 5.84E+06 1.17E+07 1.45E+07 2.41E+07 2.69E+09 O.OOE+01 O.OOE+Ol I-135 3.35E+04 5.39E+04 3.93E+04 7.07E+04 1.08E+05 6.26E+06 O.OOE+01 O.OOE+01 Cs-134 4.78E+09 1.22E+08 1.38E+ 10 2.27E+ 10 7.03E+ 09, O.OOE+ 01 2.52E+ 09 O.OOE + 01 Cs-136 1.14E+09 6.17E+07 6.39E+ 08 1.76E+ 09 9.36E + 08 O.OOE+ 01 1.40E + 08 O.OOE + 01 Cs-137 2.91E+09 1.23E+08 2.06E+ 10 1.97E+ 10 6.42E+09 O.OOE+01 2.31E+09 O.OOE+01 Ba.140 4.36E+ 06 3.78E+ 07 7.47E+ 07 6.54E+ 04 2.1 3E+ 04 O.OOE+ 01 3.90E+ 04 O.OOE+ Ol Ce-141 9.73E+02 8.17E+06 1.31E+04 6.55E+ 03 2.87E+03 O.OOE+01 O.OOE+01 O.OOE+01 Co-144 5.20E+04 7.96E+07 9.74E+05 3.05E+05 1.69E+05 O.OOE+01 O.OOE+01 O.OOE+01 Hf-181 2.53E+03 4.16E+05 2.51E+04 9.79E+01 7.86E+01 8.24E+01 O.OOE+01 O.OOE+01

'R Values in units of /nrem/yr por /ICI/m~ for inhalation and tritium and in units of mrom/yr per pCi/soc per m't for all others.

3-43

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.3-11 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'ATHWAY

~ Cow Milk AGE GROUP ~ Infant bhuJIda GLIrau Jina y XhyxaiL H-3 2.40E+03 2.40E+03 O.OOE+ 01 2.40E+ 03 2.40E+03 2.40E+03 2.40E+03 2.40E+03 P-32 4.06E+ 09 1.42E+ 09 1.05E+ 1 1 6.17E+ 09 O.OOE+ 01 O.OOE+01 O.OOE+ 01 O.OOE+ 01 Cr-51 9.77E+04 2.85E+06 O.OOE+01 O.OOE+01 1.39E+04 6.38E+04 1.24E+05 O.OOE+01 Mn-54 5.37E+06 8.71E+06 O.OOE+01 2.37E+07 5.25E+06 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Fe-59 9.23E+07 1.12E+08 1.34E+08 2.34E+08 O.OOE+01 O.OOE+ 01 6.92E+07 O.OOE+01 Co.58 3.60E+07 3.59E+07 O.OOE+0>>.44E+07 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Co.60 1.29E+08 1.30E+08 O.OOE+01 5.47E+07 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Zn-65 '.14E+09 1.12E+10 3.88E+09 1.33E+10 6.45E+09 O.OOE+01 O.OOE+01 O.OOE+ 01 Rb-86 6.86E+09 3.55E+ 08 O.OOE+ 01 1.39E+ 10 O.OOE+Ol O.OOE+01 O.OOE+01 O.OOE+ 01 Sr-89 2.17E+08 1.55E+08 7.55E+09 O.OOE+01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Sr-90 2.05E+ 10 1.00E+09 8.04E+ 10 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Y-91 1.16E+08 3.12E+06 4.36E+04 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Zr-95 7.01E+02 4.92E+05 4.05E+03 9.88E+02 1.06E+03 O.OOE+01 O.OOE+01 O.OOE+01 Nb-95 8.48E+04 1.24E+08 3.56E+05 1.47E+05 1.05E+05 O.OOE+01 O.OOE+01 O.OOE+01 Ru-103 1.74E+ 03 6.33E+ 04 5.21E+ 03 O.OOE+ 01 1.08E+ 04 O.OOE+ Ol O.OOE+ 01 O.OOE+01

'u-106 1.47E+04 8.95E+05 1.18E+05 O.OOE+01 1.39E+ 05 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Ag-110M 1.19E+08 9.32E+09 2.46E+08 1.80E+08 2.57E+08 O.OOE+01 O.OOE+01 O.OOE+ 01 Sn-113 6.65E+06 1.37E+07 6.45E+ 06 2ASE+ 05 1.31E+05 9.34E+04 O.OOE+01 O.OOE+ 01 Sb-124 3.90E+07 3.88E+08 1.26E+08 1.85E+06 O.OOE+Ol 3.34E+05 7.88E+07 O.OOE+01 Te-127M 3.75E+07 1.25E+08 3.10E+08 1.03E+08 7.64E+08 8.96E+07 O.OOE+01 O.OOE+ 01 Te-129M 5.57E+07 2.16E+08 3.62E+ 08 1.24E+08 9.05E+08 1.39E+08 O.OOE+01 O.OOE+ 01 I-'131 9.23E+08 7.49E+07 1.78E+d9 2.10E+ 09 2.45E+ 09 6.90E+ 11 O.OOE+ 01 O.OOE+Ol I-1 32 6.90E+01 1.57E.OO 9.55E+01 1.94E+00 2.16E+00 9.09E+01 O.OOE+01 O.OOE+01 1-133 1.05E+07 6.09E+ 06 2.47E+ 07 3.60E+ 07 4.23E+ 07 6.55E+ 09 O.OOE+ 01 O.OOE+ 01 I-135 5.93E+04 5.88E+ 04 8.17E+ 04 1.63E+ 05 1.81E+ 05 1.46E+ 07 O.OOE+ 01 O.OOE+01 Cs-1 34 4.19E+09 1.13E+08 2.23E+ 10 4.15E+ 10 1.07E+ 10 O.OOE+ Ol 4.38E+ 09 O.OOE+ 01 Cs-1 36 1.37E+09 5.58E+07 1.25E+09 3.67E+09 1.46E+09 O.OOE+01 2.99E+08 O.OOE+01 Cs-137 2.72E+09 1.20E+08 3.28E+10 3.84E+10 1.03E+10 O.OOE+Ol 4.18E+09 O.OOE F01 B3-140 7.91E+ 06 3.77E+ 07 1.54E+ 08 1.54E+05 3.65E p 04 O.OOE+ Ol 9.43E+ 04 O.OOE+Ol Ce-141 1.87E+03 8.21E+06 2.60E+04 1.59E+04 4.90E+03 O.OOE+Ol O.OOE+01 O.OOE+ Ol Ce-144 7.82E+04 8.01E+07 1.40E+06 5.71E+05 2.31E+05 O.OOE+01 O.OOE+Ol O.OOE+01 Hf-181 4.23E+ 03 3.93E+ 05 4.78E+ 04 2.26E+ 02 1.32E+ 02 1.91E+ 02 O.OOE+01 O.OOE+ Ol

'R Vaiues in units of mrem/yr per yCi/m'or inhalation and tritium and in units of mrem/yr per yCi/sec per m'or all others.

Shearon Harris Nuclear Power plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 I

TABLE 3.3-12 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'GE PATHWAY ~ Goat Milk GROUP ~ Adult JQdrmy IhyrnirL H-3 1.57E+03 1.576+03 O.OOE+ Ol 1.57E+03 1.576+ 03 1.57E+03 1.57E+03 1.57E+03 P-32 5.19E+ 08 1.51E+09 1.34E+ 10 8.34E+ 08 O.OOE+ 01 O.OOE+ 01 '.OOE+ 01 O.OOE+ 01 Cr-51 2.08E+03 5.23E+05 O.OOE+01 O.OOE+01 4.58E+02 1,24E+03 2.766+03 O.OOE+01 Mn.54 1.17E+05 1.88E+06 O.OOE+Ol 6.14E+05 1.83E+05 O.OOE+01 O.OOE+01 O.OOE+01 Fe-59 2.08E+05 1.81E+06 2.31E+05 5.42E+05 O.OOE+01 O.OOE+01 1.51E+05 O.OOE+01 Co.58 7.54E+ 05 6.82E+ 06 O.OOE+ 01 3.36E+ 05 O.OOE+ 01 O.OOE+ Ol 0.006+ 01 O.OOE+ 01 Co.60 2.69E+06 2.29E+07 O.OOE+01 1.22E+OS O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Zn 65 1.65E+ 08 2.31E+ 08 1.15E+ 08 3.66E+ 08 2.45E+ OS O.OOE+Ol O.OOE+ 01 O.OOE+ 01 Rb-86 9.05E+07 3.83E+07 O.OOE+01 1.94E+08 O.OOE+01 O.OOE+01 O.OOE + 01 O.OOE+ 01 Sr 89 5.24E+07 2.93E+08 1.83E+09 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+Ol Sr-90 1.59E+10 1.88E+ 09 6.496+ 10 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ Ol Y-91 1.64E+01 3.37E+05 6.136+02 0.006+01, O,OOE+ 01 O.OOE+ 01 O.OOE+ Ol O.OOE+ 01 Zr-95 1.46E+01 6.85E+04 6.74E+01 2.16E+01 3.39E+ 01 O.OOE+ 01 O.OOE+01 O.OOE+01 Nb.95 1.78E+03 2.01E+ 07 5.94E+ 03 3.31E+ 03 3.27E+ 03 O.OOE+01 O.OOE+Ol O.OOE+01 Ru-103 3.1 6E+ 01 8.56E+ 03 7.33E+ 01 O.OOE+ 01 2.806+ 02 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 RU-106 1.92E+02 9.81E+04 1.52E+03 0.006+01 2.93E+ 03 O.OOE+01 O.OOE+01 O.OOE+01 Ag-110M 2:45E+06 1.68E+09 4.46E+06 4.12E+06 8.11E+06 O.OOE+01 O.OOE+01 O.OOE+01 Sn-113 '.326+05 2.44E+06 1.40E+05 5,41E+03 3.96E+03 1.90E+03 O.OOE+01 O.OOE+01 Sb-124 7.36E+05 5.27E+07 1.86E+06 3.51E+04 O.OOE+Ol 4.50E+03 1.44E+06 O.OOE+01 Te-127M 4.93E+ 05 1.36E+ 07 4.05E+ 06 1.45E+ 06 1.64E+07 1.03E p06 O.OOE+ 01 O.OOE+ 01 Te-129M 7.43E+05 2.36E+07 4.69E+06 1.75E+06 1.96E+07 1.61E+06 0.006+01 O.OOE+Ol l-131 1.91E+08 8.78E+07 2.33E+ 08 3.33E+08 5.71E+ 08 " 1.09E+ 11 O.OOE+ 01 O.OOE+ Ol I-1 32 1.23E+01 6.61E-02 1.32E+01 3.52E+01 5.61E+Ol 1.23E+01 O.OOE+01 O.OOE+01 I ~ 133 1.68E+ 06 4.95E+ 06 3.17E+06 5.51E+06 9.61E+ 06 8.10E+ 08 O.OOE+01 O.OOE+ 01 I-135 1.08E+04 3.32E+04 1.126+04 2.94E+04 4.71E+04 1.94E+06 O.OOE+01 O.OOE+01 Cs-1 34 2.01E+10 4.31E+08 .1.03E+ 10 2.46E+ 10 7.97E+ 09 O.OOE+ 01 2.65E+ 09 O.OOE+ 01 Cs-1 36 1.42E+09 2.24E+08 4.99E+ 08 1.97E+09 1.106+09 O.OOE+01 1.50E+08 O.OOE+01 Cs-137 1.27E+ 10 3.74E+ 08 1.41E+ 10 1.936+10 6.56E+09 O.OOE+01 2.18E+09 0.006+01 Ba-140 1.35E+05 4.23E+06 2.06E+ 06 2.58E + 03 8.78E+ 02 O.OOE+ 01 1.48E + 03 O.OOE + 01 Ce-141 2.686+ Ol 9.036+05 3A9E+ 02 2.36E+ 02 1.10E+ 02 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Ce-144 1.38E+03 8.71E+06 2.58E+ 04 1.08E+ 04 6.39E + 03 O.OOE+ Ol O.OOE + 01 O.OOE + 01 Hf-181 8.02E+01 5.26E+04 7.09E+02 3.99E+00 3.34E+00 2.54E+00 O.OOE+01 0.006+01

'R Values in units of mrcm/yr pcr pCi/mc for inhalation and tritium and in units of mrem/yr per pCi/scc per m'or all others.

3-45

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.3-13 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'GE PATHWAY ~ Goat Milk GROUP = Teen JCuhtrty XhyLoitL H-3 2.04E+03 2.04E+03 O.OOE+01 2.04E+03 2.04E+ 03 2.04E+ 03 2.04E+ 03 2.04E+ 03 P-32 9.60E+ 08 2.08E+ 09 2.48E+ 10 1.53E+09 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Cr-51 3.63E+03 6.10E+05 O.OOE+Ol O.OOE+01 7.95E+02 2.02E+03 5.18E+03 O.OOE+01 Mn.54 2.03E+ 05 2.10E+ 06 O.OOE+ 01 1.02E+ 06 3.05E+ 05 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Fe-59 3.63E+05 2.22E+06 4.03E+05 9.40E+05 O.OOE+ 01 O.OOE + 01 2.96E + 05 O.OOE + 01 Co-58 1.30E+ 06 7.80E+ 06 O.OOE+ 01 5.66E+ 05 O.OOE+ Ol O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Co 60 4.66E+06 2.69E+07 O.OOE+01 2.07E+06 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Zn-65 . 2.86E+08 2.60E+08 1.77E+08 6.13E+08 3.93E+ 08 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Rb-86 1.66E+08 5.24E+07 O.OOE+Ol 3.54E+08 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+Ol Sr-89 9.65E+07 4.01E+08 3.37E+09 O.OOE+01 O.OOE +01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Sr-90 2.27E+10 2.58E+09 9.18E+ 10 O.OOE+01 O.OOE+01 O.OOE+Ol O.OOE+01 O.OOE+ 01 Y-91 3.02E+ 01 4.62E+ 05 1.1 3E+ 03 O.OOE+ 01 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Zr-95 2.56E+01 8.59E+04 1.18E+02 3.72E+01 5.47E+01 O.OOE+01 O.OOE+01 O.OOE+ 01 Nb-95 3.09E+03 2.40E+07 1.01E+04 5.62E+03 5.45E+03 O.OOE+01 O.OOE+Ol O.OOE+01 RU-103 5.58E+01 1.09E+04 1.30E+02 O.OOE+01 4.60E+ 02 O.OOE+Ol O.OOE+01 O.OOE+01

\

Ru-106 3.51E+02 1.34E+05 2.79E+ 03 O.OOE+ 01 5.38E+ 03 O.OOE+01 O.OOE+01 O.OOE+Ol Ag-110M 4.24E+06 1.96E+09 7.37E+ 06 6.97E+ 06 1.33E+ 07 O.OOE+ 01 O.OOE+ 01 O.OOE+01 Sn-113 2.28E+05 2.58E+06 2.15E+05 9.06E+03 6.37E+03 2.97E+03 O.OOE+ 01 O.OOH. 01 Sb-124 1.29E+06 6.67E+07 3.31E+06 6.10E+04 O.OOE+01 7.51E+03 2.89E+ 06 O.OOE+ 01 Te-127M 8.87E+05 1.86E+07 7.46E+06 2.65E+06 3.02E+ 07 1.77E+06 O.OOE+01 O.OOE+ 01 Te-129M 1.36E+06 3.22E+07 8.58E+06 3.19E+06 3.59E+ 07 2.77E+06 O.OOE+01 O.OOE+01 I-131 3.18E+08 1.17E+08 4.22E+ 08 5.91E+ 08 1.02E+ 09 1.73E+ 11 O.OOE+ 01 O.OOE+01 I-132 2.19E+Ol 2.66E+01 2.33E+ 01 6.1 1E+ 01 9.62E+ 01 2.06E+ 01 O.OOE+01 O.OOE+ 01 I-133 2.99E+ 06 7.43E+ 06 5.79E+06 9.81E+06 1.72E+07 1.37E+09 O.OOE+01 O.OOE+01 1-135 1.90E+ 04 5.63E+ 04 1.99E+04 5.13E+04 8.10E+04 3.30E+06 O.OOE+01 O.OOE+01 Cs-1 34 1.96E+ 10 5.26E+08 1.80E+ 10 4.23E+ 10 1.34E+ 10 O.OOE+ 01 5.13E+ 09 O.OOE+ 01 Cs-1 36 2.25E+09 2.69E+07 8.50E+ 08 3.34E+09 1.82E+09 O.OOE+01 2.87E+08 O.OOE+01 Cs-1 37 1.19E+10 4.85E+08 2.56E+10 3.41E+10 1.16E+10 O.OOE+Ol 4.51E+09 O.OOE+01 Ba-140 2.39E+ 05 5.72E+ 06 3.71E+ 06 4.55E+ 03 1.54E+ 03 O.OOE+ 01 3.06E+ 03 O.OOE+ 01 Ce-141 4.91E+ 01 1.22E+06 6.40E + 02 4.27E+ 02 2.01E+ 02 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Ce-144 2.55E+03 1.19E+07 4.74E+04 1.96E+04 1.17E+04 O.OOE+ 01 O.OOE+ 01 O.OOE + 01 Hf-181 1.41E+ 02 6.34E+ 04 1.27E+ 03 6.97E+ 00 5.80E+ 00 4.26E+ 00 O.OOE+ 01 O.OOE+ 01 R Veiues in units of mrem/yr per yCI/m~ for inhalation end tritium end in units of mrem/yr per yCi/sec per m'or eli others.

3-46

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.3-14 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANTS PATHWAY ~ Goat Milk AGE GROUP ~ Child LBodg JQdnay Ihyraid H-3 3.23E+03 3.23E+03 O.OOE+01 3.23E+03 3.23E+03 3.23E+03 3.23E+03 3.23E+03 P-32 2.35E+09 1.69E+09 6.11E+10 2.86E+09 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Cr.51 7 40E+03 3.93E+05 O.OOE+01 O.OOE+Ol 1.12E+03 4.11E+03 7.50E+03 O.OOE+01 Mn-54 4.07E+ 05 1.28E+ 06 O.OOE+ 01 1.53E+06 4.29E+05 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Fe-59 7.52E+05 1.57E+06 9.34E+05 1.51E+06 O.OOE+01 O.OOE+01 4.38E+05 O.OOE+01 Co.58 2.65E+06 5.05E+06 O.005+ 01 8,65E+05 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Co-60 9.48E+06 1.78E+07 O.OOE+01 3.21E+06 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Zn-65 5.74E+08 1.62E+08 3A7E+08 9.24E+08 5.82E+08 O.OOE+01 O.OOE+01 O.OOE+01 Rb-86 4.04E+ 08 4.22E+ 07 O.OOE+ 01 6.57E+ 08 O.OOE+ 01 O.OOE+ 0) O.OOE+ 01 O.OOE+ 01 Sr-89 2.38E+08 3.23E+08 8.34E+09 O.OOE+01 O.OOE+Ol O.OOE+ 01 O.OOE+01 O.OOE+01 Sr-90 3.93E+10 2.09E+09 1.55E+11 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+Ol Y-91 7.45E+01 3.71E+05 2.79E+03 O.OOE+Ol O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Zr-95 5.36E+Ol 6.28E+04 2.74E+02 6.02E+01 8.62E+01 O.OOE+Ol O.OOE+01 O.OOE+01 Nb-95 6.37E+03 1.65E+07 2.29E+04 8.91E+ 03 8.37E+ 03 O.OOE+ Ol O.OOE+ 01 O.OOE+ 01 Ru-103 1.19E+02 7.98E+03 3.09E+02 O.OOE+Ol 7.77E+02 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Ru-106 '.56E+02 1.07E+05 6.86E+ 03 O.OOE + 01 9.27E+03 O.OOE+Ol O.OOE+01 O.OOE+01 Ag-110M 8.63E+ 06 1.28E+ 09 1.60E+07 1.08E+ 07 2.01E+ 07 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Sn-113 4.61E+ 05 1.69E+06 4.22E+05 1.3'6E+ 04 9.38E+ 03 5.59E+ 03 O.OOE+01 O.OOE+01 Sb-124 2.75E+ 06 4.91E+07 7.84E+ 06 1.02E+ 05 O.OOE+ 01 '.73E+ 04 4.35E+ 06 O.OOE+ 01 Te-127M'.18E+06 1.49E+07 1.84E+ 07 4.95E+06 5.24E+07 4.40E+06 O.OOE+01 O.OOE+01 Te-129M 3.28E+06 2.58E+07 2.12E+ 07 5.91E+06 6.21E+07 6.82E+06 O.OOE+01 O.OOE+01 l-131 5.85E+ 08 9.17E+07 1.02E+09 1.03E+ 09 1.69E+ 09 3.41E+ 11 O.OOE+ 01 O.OOE+ 01 I-132 4.67E+ 01 1.19E+00 5.52E+ 01 1.01E+ 00 1.55E+ 00 4.71E+01 O.OOE+ 01 O.OOE+ 01 I-133 6.58E+ 06 7.00E+ 06 1.41E+07 1.74E+ 07 2.90E+ 07 3.23E+ 09 O.OOE+ 01 O.OOE+ Oi I-135 4.01E+ 04 6.47E+04 4.72E+04 8.49E+ 04 1.30E+ 05 7.52E+ 06 O.OOE+ 01 O.OOE+01 Cs-134 1.43E+10 3.67E+08 4.14E+10 6.80E+ 10 2.11E+ 10 O.OOE+ 01 7.56E+ 09 O.OOE+ 01 Cs-136 3.41E+09 1.85E+08 1.92E+09 5.27E+09 2.81E+09 O.OOE+01 4.19E+08 O.OOE+01 Cs-1 37 8.72E+ 09 3.70E+ 08 6.17E+ 10 5.91E+ 10 1.93E+ 10 O.OOE+ 01 6.93E+ 09 O.OOE+ Ol Ba-140 5.23E+05 4.54E+05 8.96E+06 7.85E+03 2.56E+03 O.OOE+01 4.68E+03 O.OOE+01 Ce-141 1.17E+02 9.81E+05 1.53E+03 7.36E+02 3.45E+ 02 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Ce-144 6.24E+03 9.55E+06 1.17E+05 3.66E+04 2.03E+ 04 O.OOE+ Ol O.OOE+ Ol O.OOE + Ol Hf-181 3.04E+ 02 4.99E+ 04 3.02E+ 03 1.17E+ 01 9.43E+ 00 9.88E+ 00 O.OOE+ 01 O.OOE+ 01

'R Values in units of mrem/yr per yCi/m'or inhalation and tritium and in units of mrcm/yr per yCi/sec per ma for all others.

3-47

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TA8LE 3.3-15 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'GE PATHWAY ~ Goat Milk GROUP ~ Infant XJhuiy. Xiidttrty H-3 4.90E+03 4.90E+03 O.OOE+ 01 4.90E+ 03 4.90E+ 03 4.90E+ 03 4.90E+03 4.90E+ 03 P-32 4.88E+ 09 1.70E+ 09 1.26E+ 11 7.40E+ 09 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ Ol Cr-51 1.17E+ 04 3.42E+ 05 O.OOE+ Ol O.OOE+ 01 1.67E+ 03 7.65E+ 03 1.49E+ 04 O.OOE+ 01 Mn-54 6.45E+ 05 1.04E+ 06 O.OOE+01 2.84E+06 6.30E+05 O.OOE+01 O.OOE+01 O.OOE+01 FQ-59 1.20E+06 1.45E+06 1.74E+06 3.04E+06 O.OOE+01 O.OOE+01 9.00E+05 O.OOE+01 Co-58 4.31E+06 4.31E+06 O.OOE+Ol 1.73E+06 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Co-60 1.55E+07 1.56E+07 O.OOE+01 6.56E+06 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Zn-65 ~ 7.36E+ 08 1.35E+ 09 4.66E+ 08 1.60E+ 09 7.74E+ 08 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Rb-86 8.23E+08 4.26E+07 O.OOE+01 1.67E+09 O.OOE+Ol O.OOE+01 O.OOE+01 O.OOE+01 Sr-89 4.55E+ 08 3.26E+ 08 1.59E+ 10 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O,OOE+ 01 O.OOE+ 01 Sr-90 4.30E+ 10 2.11E+ 09 1.69E+ 1 1 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+01 O.OOE+01 Y-91 1.39E+ 02 3.75E+05 5.23E+ 03 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ Ol O.OOE+ 01 Zr-95 8.41E+ 01 5.90E+04 4.85E+02 1.19E+02 1.28E+02 O.OOE+01 O.OOE+01 O.OOE+01 Nb.95 1.02E+04 1.48E+07 4.27E+04 1.76E+04 1.26E+04 O.OOE+01 O.OOE+01 O.OOE+01 Ru-103 2.09E+02 7.80E+03 6.25E+02 O.OOE+01 1.30E+03 O.OOE+01 O.OOE+01 O.OOE+01 Ru-106 1.77E+03 1.07E+05 1.41E+04 O.OOE+01 1.67E+04 O.OOE+01 O.OOE+01 O.OOE+Ol Ag-110M 1143E+07 1.12E+09 2.95E+07 2.16E+07 3.08E+07 O.OOE+01 O.OOE+01 O.OOE+01 Sn-113 6.66E+05 1.37E+06 6.46E+05 2.45E+04 1.32E+04 9.34E+03 O.OOE+01 O.OOE+ 01 Sb-124 4.68E+06 4.66E+07 1.51E+07 2.22E+05 O.OOE+01 4.01E+ 04 9.46E+ 06 O.OOE+ 01 Te-127M 4.51E+ 06 1.50E+07 3.72E+07 1.23E+07 9.16E+07 1.08E+07 O.OOE+ 01 O.OOE+ 01 Te-129M 6.69E+ 06 2.59E+07 4.34E+ 07 1.49E+07 1.09E+08 1.67E+07 O.OOE+01 O.OOE+01 I-131 1.1 1E+ 09 8.99E+ 07 2.14E+09 2.52E+ 09 2.94E+ 09 8.28E+ 11 O.OOE+ 01 O.OOE+ 01 1-132 8.28E+ 01 1.88E+00 1.15E+00 2.33E+00 2.59E+00 1.09E+02 O.OOE+Ol O.OOE+01 I-133 1.27E+ 07 7.31E+06 2.97E+07 4.32E+07 5.08E+07 7.86E 4 09 O.OOE+ 01 O.OOE+ 01 I-135 7.11E+04 7.06E+04 9.81E+04 1.95E+05 2.17E+05 1.75E+07 O.OOE+01 O.OOE+01 Cs-134 1.26E+10 3.38E+ 08 6.68E+ 10 1.25E+ 11 3.21E+ 10 O.OOE+ 01 1.31E+ 10 O.OOE+ 01 Cs-136 4.11E+09 1.67E+08 3.75E+ 09 1.10E+10 4.39E+09 O.OOE+01 8.98E+08 O.OOE+01 Cs-1 37 8.17E+ 09 3.61E+ 08 9.85E+ 10 1.15E+ 11 3.10E p 10 O.OOE+01 1.25E+10 O.OOE+01 Ba-140 9.50E+05 4.53E+06 1.84E+07 1.84E+04 "

4.38E+ 03 O.OOE+ 01 1.13E+ 04 O.OOE+ 01 C0-141 2.24E+02 9.85E+05 3.13E+03 1.91E+03 5.88E+02 O.OOE+01 O.OOE+01 O.OOE+ 01 Ce-144 9.39E+03 9.61E+06 1.67E+05 6.86E+04 2.77E+04 O.OOE+Ol O.OOE+ Ol O.OOE+ 01 Hf-181 5.08E+02 4.72E+04 5.74E+ 03 2.71E+01 1.58E+01 2.30E+01 O.OOE+01 O.OOE+01

'R VaIues in units of mrem/yr per yCi/m'or inhalation and tritium and in units of mrcm/yr per yCi/sec per m'or all others.

3-48

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.3-16 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'GE PATHWAY ~ Inhal GROUP ~ Adult LJhdy. QLIracI edna@ Ihyuud H-3 1.26E+03 1.26E+03 O.OOE+ 01 1.26E+ 03 1.26E+03 1.26E+03 1.26E+03 1.26E+03 P-32 5.00E+04 8.63E+04 1.32E+06 7.70E+04 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 Cr-51 9.99E+ Ol 3.32E+ 03 O.OOE+ 01 O.OOE+ 01 2.28E+01 5.94E+01 1.44E+04 O.OOE+01 Mn.54 6.29E+03 7.72E+04 O.OOE+01 3.95E+04 9.83E+03 O.OOE+01 1.40E+06 O.OOE+01 Fe.59 1.05E+04 1.88E+05 1.17E+04 2.77E+04 O.OOE+Ol O.OOE+01 1.01E+06 O.OOE+01 Co-58 2.07E+03 1.06E+05 O.OOE+01 1.58E+03 O.OOE+ 01 O.OOE+ 01 9.27E+05 O.OOE+ 01 Co-60 1.48E+04 2.84E+05 O.OOE+Ol 1.15E+04 O.OOE+ 01 O.OOE+ 01 5.96E+06 O.OOE+01 Zn-65 ~

4.65E+ 04 5.34E+ 04 3.24E+ 04 1.03E+ OS 6.89E+ 04 O.OOE+ 01 8.63E+ 05 O.OOE+ 01 Rb.86 5.89E+04 1.66E+04 O.OOE+01 1.35E+05 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+Ol Sr-89 8.71E+03 3.49E+05 3.04E+05 O.OOE+01 O.OOE+ 01 O.OOE+ Ol 1.40E + 06 O.OOE+ 01 Sr-90 6.09E+06 '.21E+05 9.91E+07 O.OOE+01 O.OOE +01 O.OOE+ 01 9.59E+ 06 O.OOE+01 Y-91 1.24E+ 04 3.84E+ 05 4.62E+05 O.OOE+ Ol O.OOE+ Ol O.OOE+ 01 1.70E+ 06 O.OOE+01 Zr-95 2.32E+04 1.50E+05 1.07E+05 3.44E+04 5.41E+04 O.OOE+01 1.77E+06 O.OOE+01 Nb-95 4.20E+03 1.04E+05 1.41E+04 7.80E+03 7.72E+ 03 O.OOE+01 5.04E+05 O.OOE+ 01 Ru-103 6.57E+02 1.10E+05 1.53E+03 O.OOE+01 5.82E+ 03 O.OOE+01 5.04E+05 O.OOE+ 01 Ru-106 8.71E+03 9.11E+05 6.90E+ 04 O.OOE+ 01 1.33E+05 O.OOE+01 9.35E+06 O.OOE+01 Ag-110M 5.94E+03 3.02E+05 1.08E+04. 9.99E+03 1.97E+04 O.OOE+01 4.63E+06 O.OOE+01 Sn-113 6.48E+ 03 2.48E+ 04 6.86E+03 2:69E+02 1.97E+02 9.33E+ 01 2.99E+ 05 O.OOE+ 01 Sb-124 1.24E+ 04 4.06E+ 05 3.12E+04 5.88E+02 O.OOE+01 7.55E+ 01 2.48E+06 O.OOE+01 Te-127M 1.57E+03 1.49E+05 1.26E+ 04 5.76E+ 03 4.57E+04 3.28E+03 9.59E+ 05 O.OOE+ 01 Te-129M 1.58E+ 03 3.83E+ 05 9.75E+03 4.67E+03 3.65E+04 3.44E+03 1.16E+06 O.OOE+Ol I-131 2.05E+04 6.27E+03 2.52E+04 3.57E+04 6.12E+04 1.19E+07 O.OOE+ 01 O.OOE+ 01 I-132 1.16E+03 4.06E+02 1.16E+03 3.25E+03 5.18E+ 03 1.14E+05 O.OOE+01 O.OOE+ 01 1-133 4.51E+03 8.87E+03 8.63E+03 1.48E+04 2.58E+04 2.15E+06 O.OOE+ 01 O.OOE+ 01 I ~ 135 2.S6E+03 5.24E+03 2.68E+03 6.97E+03 1.11E+04 4.47E+ 05 O.OOE+ 01 O.OOE+ 01 Cs-134 7.27E+ 05 1.04E+ 04 3.72E+ OS 8.47E+ 05 2.87E+ 05 O.OOE+ 01 9.75E+04 O.OOE+01 Cs-136 1.10E+05 1.17E+04 3.90E+04 1.46E+05 8.55E+04 O.OOE+ 01 1.20E+ 04 O.OOE+ 01 Cs-1 37 4.27E+05 8.39E+03 4.78E+05 6.20E+05 2.22E+05 O.OOE+01 7.51E+04 O.OOE+01 Ba-140 2.56E+03 2.18E+05 3.9OE p04 4.90E+01 1.67E+ 01 O.OOE+ 01 1.27E+ 06 O.OOE+ 01 Ce-141 1.53E+03 1.20E+05 1.99E+04 1.35E+ 04 6.25E+ 03 O.OOE+ 01 3.61E+ 05 O.OOE+01 Ce.144 1.84E+05 8.15E+05 3.43E+06 1.43E+06 8.47E+ 05 O.OOE+ 01 7.76E+ 06 O.OOE+ 01 Hf-181 5.16E+03 1.29E+05 4.56E+04 2.57E+02 2.15E+02 1.63E+02 5.99E+05 O.OOE+01 I

'R Values in units of mrem/yr per yCI/m~ for inhalation and tritium and in units of mrem/yr per yCI/sec per m't for all others.

3-49

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.3-17 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'GE PATHWAY ~ Inhai GROUP ~ Teen

'B'LXrau JCufnay IhyrntL H-3 1.27E+03 1.27E+03 O.OOE+ 01 1.27E+ 03 1.27E+03 1.27E+ 03 1.27E+03 1.27E+03 P-32 7.15E+04 9.27E+04 1.89E+06 1.09E+05 O.OOE+01 O.OOE+Ol O.OOE+01 O.OQE+Ol Cr-51 1.35E+02 3.0QE+03 O.OOE+01 O.OOE+01 3.07E+01 7.49E+01 2.09E+04 O.OOE+01 Mn-54 8.39E+03 6.67E+04 O.OQE+ Ol 5.10E+ 04 1.27E+ 04 O.OOE+ 01 1.98E+ 06 O.OOE+ 01 Fe-59 1 43E+04 1.78E+05 1.59E+ 04 3.69E+04 O.OOE+ Ol O.OOE+ 01 1.53E+ 06 O.QQE+ 01 Co-58 2.77E+03 9.51E+04 O.OOE+01 2.07E+03 O.OOE+01 O.OOE+01 1.34E+06 O.OOE+ 01 Co.60 1.98E+ 04 2.59E+ 05 O.OOE+ 01 1.51E+ 04 O.OQE+ 01 O.OOE+ 01 8.71E+ 06 O.OOE+ 01 Zn.65 6.23E+ 04 4.66E+ 04 3.85E+04 1.33E+05 8.63E+04 O.OOE+ Ol 1.24E+06 O.OOE+ 01 Rb-86 8.39E+04 1.77E+04 O.OQE+ Q 1 1.9QE+ 05 O.OOE+ 01 O.OOE+Ol O.OOE+Ol O.OQE+Ol Sr-89 1.25E+04 3.71E+05 4.34E+ 05 O.OOE+ 01 O.OOE+ 01 O.OOE+01 2.41E+06 O.OOE+01

\

Sr-90 6.67E+ 06 7.64E+ 05 1.08E+08 O.OOE+01 O.OOE+ 01 O.OOE+01 1 65E+07 0 OQE+01 Y-91 1.77E+04 4.08E+ 05 6.60E+05 O.OQE+ 01 O.OOE+01 O.OOE+Ol 2.93E+06 O.OOE+01 Zr-95 3.15E+ 04 1.49E+05 1.45E+05 4.58E+04 6.73E+04 O.OOE+Ol 2.68E+06 O.QQE+Ol Nb-95 5.66E+ 03 9.67E+04 1.85E+04 1.03E+04 9.99E+03 O.OOE+01 7.50E+ 05 O.OOE+ 01 RU-103 8.95E+ 02 1.09E+05 2.10E+03 O.OOE + 01 7.42E+ 03 O.OOE+ 01 7.82E+05 O.OOE+ 01 Ru.106 1.24E+ 04 9.59E+ 05 9.83E+ 04 O.OOE+ 01 1.90E+05 O.OOE+ 01 1 61E+07 0 OOE+01 Ag-110M 7.98E+ 03 2.72E+ 05 1.38E+04 1.31E+04 2.50E+04 O.OOE+01 6.74E+ 06 O.OQE+ 01 Sn-113 8.68E+ 03 2.03E+ 04 8.19E+03 9,44E+02 2.45E+02 1.13E+02 4.27E+05 O.OQE+01 Sb-124 1.68E+ 04 3.98E+ 05 4.30E+ 04 7.94E+02 O.OOE+01 9.76E+01 3.85E+06 O.OQE+01 Te-127M 2.18E+03 1.59E+05 1.80E+04 8.15E+03 6.53E+04 4.38E+03 1.65E+06 O.OQE+01 Te-129M 2.24E+03 4.04E+05 1.39E+04 6.57E+03 5.18E+04 4.57E+03 1.97E+06 O.OOE+Ol I-131 2.64E+ 04 6.48E+ 03 3.54E+ 04 4.90E+ 04 8.39E+04 1.46E+07 O.OQE+Ol O.OQE+Ol I-132 1.57E+03 1.27E+03 1.59E+03 4.37E+ 03 6.91E+03 1.51E+05 O.OOE+01 O.OQE+ 01 I-133 6.21E+03 1.03E+04 1.21E+04 2.05E+04 3.59E+04 2.92E+06 O.OQE+01 O.OOE+Ol I-1 35 3.48E+03 6.94E+03 3.69E+03 9.43E+03 1.49E+ 04 6.20E+ 05 O.OOE+ 01 O.OOE+ 01 Cs-134 5.48E+ 05 9.75E+03 5.02E+05 1.13E+06 3.75E+ 05 Q.OOE+ Ol 1.46E+05 O.OOE+ Ol Cs-136 1.37E+ 05 1.09E+ 04 5.14E+04 1.93E+05 1.10E+ 05 O.OOE+ 01 1.77E+ 04 O.OOE+ 01 Cs-137 3.1 1E+ 05 8.48E+ 03 6.69E+ 05 8.47E+05 3.04E+ 05 O.OOE+ 01 1.21E+05 O.OOE+ 01 Ba-140 3.51E+03 2.28E+ 05 5.46E+ 04 6.69E+Ol 2.28E+ 01 O.OQE+ 01 2.03E+ 06 O.QOE+Ol Ce-141 2.16E+03 1.26E+05 2.84E+ 04 1.89E+ 04 8.87E+ 03 O.OOE+ Ol 6.13E+05 O.QOE+Ol Ce-144 2.62E+ 05 8.63E+ 05 4.88E+ 06 2.02E+06 1.21E+06 O.OOE+ 01 1.33E+ 07 O.OOE + 01 Hf-181 7.04E+03 1.20E+05 6.32E+04 3.48E+02 2.9OE+02 2.12E+02 9.39E+05 O.OQE+ 01

'R Values in units of mremlyr per lrCilm'or inhaiation and tritium and in units of mremlyr per pCilseo per m'~ for all others.

3-50

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.3-1 8 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'ATHWAY

~ Inhai AGE GROUP ~ Child Kidney Z~roid H-3 1.12E+03 1.12E+03 O.OOE+ 01 1.12E+03 1.1 2E+ 03 1.12E+03 1.12E+03 1.12E+03 P-32 9.86E+04 4.21E+04 2.60E+06 1.14E+05 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Cr-51 1.54E+02 1.08E+03 O.OOE+01 O.OOE+Ol 2A3E+01 8.53E+01 1.70E+04 O.OOE+01 Mn.54 9.50E+03 2.29E+04 O.OOE+Ol 4.29E+04 1.00E+04 O.OOE+Ol 1.57E+06 O.OOE+01 Fe-59 1.67E+04 7.06E+04 2.07E+04 3.34E+04 O.OOE+01 O.OOE+01 1.27E+06 O.OOE+01 Co-58 3.16E+03 3.43E+04 O.OOE+Ol 1.77E+03 O.OOE+01 O.OOE+01 1.10E+06 O.OOE+01 Co-60 2.26E+ 04 9.61E+ 04 O.OOE+ 01 1.31E+04 O.OOE+ 01 O.OOE+Ol 7.06E+ 06 O.OOE+ 01 Zn-65 7.02E+ 04 1.63E+04 4.25E+04 1.13E+05 7.13E+04 O.OOE+01 9.94E+05 O.OOE+01 Rb.86 1.14E+05 7.98E+03 O.OOE+01 1.98E+05 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Sr-89 1.72E+04 1.67E+05 5.99E+05 O.OOE+Ol O.OOE+01 O.OOE+01 2.15E+06 O.OOE+01 Sr-90 6.43E+ 06 3.43E+ 05 1.01E+ 08 O.OOE+01 O.OOE+ 01 O.OOE+ 01 1.47E+ 07 O.OOE+ 01 Y-91 2.43E+04 1.84E+ 05 9.13E+ 05 O.OOE+ 01 O.OOE+ 01 O.OOE+ 01 2.62E+ 06 O.OOE+ 01 Zr-95 3.69E+ 04 6.10E+04 1.90E+05 4.17E+04 5.95E+04 O.OOE+Ol 2.23E+06 O.OOE+01 Nb-95 6.54E+03 3.69E+04 2.35E+ 04 9.16E+03 8.61E+03 O.OOE+01 6.13E+05 O.OOE+01 Ru-103 1.07E+ 03 4.47E+ 04 2.79E+ 03 O.OOE+01 7.02E+ 03 O.OOE+01 6.61E+05 O.OOE+01

\

Ru-106 1.69E+ 04 4.29E+ 05 1.36E+ 05 O.OOE+ 01 1.84E+05 O.OOE+ 01 1.43E+07 O.OOE+ 01 Ag-110M 9:1 3E+ 03 1.00E+05 1.68E+04 1.14E+04 2.12E+04 O.OOE+ 01 5.47E+ 06 O.OOE+01 Sn-113 9.83E+ 03 7.45E+ 03 9.00E+ 03 2.91E+02 2.02E+02 1.19E+02 3.40E+05 O.OOEw01 Sb-124 2.00E+04 1.64E+ 05 5.73E+ 04 7.40E+ 02 O.OOE+01 1.26E+02 3.24E+06 O.OOE+ 01, Te-127M 3.01E+03 7.13E+04 2.48E+04 8.53E+03 6.35E+04 6.06E+03 1.48E+06 O.OOE+ 01 Te-129M 3.04E+ 03 1.81E+ 05 1.92E+04 6.84E+03 5.02E+04 6.32E+03 1.76E+06 O.OOE+01 1-131 2.72E+04 2.84E+03 4.80E+Q4 4.80E+04 7.87E+ 04 1.62E+07 O.OOE+01 O.OOE+01 I-132 1.87E+ 03 3.20E+ 03 2.11E+03 4.06E+ 03 6.24E+ 03 1.93E+ 05 O.OOE+ 01 O.OOE+ 01 I-133 7.68E+ 03 5.47E+ 03 1.66E+ 04 2.03E+ 04 3.37E+ 04 3.84E+ 06 O.OOE+ 01 O.OOE+ 01 I-135 4.14E+03 4.43E+03 4.91E+03 8.72E+03 1.34E+04 7.91E+05 O.OOE+01 O.OOE+Ol Cs-134 2.24E+ 05 3.84E+ 03 6.50E+ 05 1.01E+06 3.30E+05 O.OOE+01 1.21E+05 O.OOE+01 Cs-136 1.16E+05 4.17E+03 6.50E+ 04 1.71E+05 9.53E+04 O.OOE+01 1.45E+04 O.OOE+01 Cs-137 1.28E+05 3.61E+03 9.05E+05 8.24E+05 2.82E+05 O.OOE+01 1.04E+05 O.OOE+Ol Ba-140 4.32E+03 1.02E+05 7.39E+04 6.47E+01 2.11E+01 O.OOE+01 1.74E+06 O.OOE+ 01 Ce-141 2.89E+03 5.65E+04 3.92E+04 1.95E+04 8.53E+03 O.OOE+ 01 5.43E+ 05 O.OOE+ Ol Ce-144 3.61E+ 05 3.88E+ 05 6.76E+ 06 2.11E+ 06 1.17E+06 O.OOE+ 01 1.19E+07 O.OOE+01 Hf-181 8.50E+03 5.31E+04 8A4E+04 3.28E+02 2.64E+ 02 2.76E+02 7.95E+05 O.OOE+01

'R Values in units of mrem/yr per /rCI/m~ for inhaIation and tritium and in units of mrern/yr per yCI/sec per m'or all others.

3-51

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 TABLE 3.3-19 R VALUES FOR THE SHEARON HARRIS NUCLEAR POWER PLANT'GE PATHWAY ~ Inhal GROUP = Infant QLXcau JCidnny Ilttrrnid H.3 6.46E+ 02 6.46E+02 O.OOE+ 01 6.46E+ 02 6.46E+ 02 6.46E+ 02 6.46f +02 6.46E+ 02 P-32 7.73E+04 1.61E+04 2.03E+06 1.12E+05 O.OOE+Ol O.OOE+01 O.OOE+01 O.OOE+01 Cr-51 8.93E+ 01 3.56E+ 02 O.OOE+ 01 O.OOE+ 01 1.32E+ 01 5.75E+ 01 1.28E+04 O.OOE+01 Mn-54 4.98E+03 7.05E+03 O.OOE+ 01 2.53f + 04 4.98E+ 03 O.OOE+ 01 9.98E+ 05 O.OOE+01 Fe.59 9.46E+03 2.47E+04 1.35E+04 2.35E+04 O.OOE+01 O.OOE+01 1.01E+06 O.OOE y 01 Co-58 1.82E+03 1.11E+04 O.OOE+01 1.22E+03 O.OOE+ 01 O.OOE+01 7.76E+05 O.OOE+ 01 Co-60 1.1 8E+ 04 3.1 9E+ 04 O.OOE+ 01 8.01E+ 03 O.OOE+ 01 O.OOE+Ol 4.50E+06 O.OOE+01 Zn-65 3.10E+04 5.13E+04 1.93E+04 6.25E+04 3.24E+ 04 O.OOE+ 01 6.46E+ 05 0.00f + 01 Rb-86 8.81E+04 3.03E+03 O.OOE+ 01 1.90E+05 O.OOE+01 O.OOE+01 O.OOE+01 O.OOE+01 Sr-89 1.14E+04 6.39E+04 3.97E+05 O.OOE+01 O.OOE+Ol O.OOE+01 2.03E+06 O.OOE+01 Sr.90 2.59E+ 06 1.31E+ 05 4.08E+ 07 O.OOE+01 O.OOE+01 0.00f + 01 1.12E+ 07 O.OOE+ 01 Y-91 1.57E+04 7.02E+04 5.87E+05 O.OOE+01 O.OOE+01 O.OOE+01 2.45E+06 O.OOE+Ol Zr.95 2.03E+04 2.17E+04 1.15E+05 2.78E+04 3.10E+04 O.OOE+01 1.75E+06 O.OOE+Ol Nb-95 3.77E+ 03 1.27E+04 1.57E+04 6.42E+03 4.71E+03 O.OOE+01 4.78E+05 O.OOE+01 Ru-103 6.78E+ 02 1.61E+04 2.01E+03 O.OOE+01 4.24E+ 03 O.OOE+ 01 '.51 E+ 05 O.OOE+ 01 I

Ru-106 1.09E+04 1.64E+05 8.67E+04 O.OOE+Ol 1:06E+05 O.OOE+ 01 1.15E+ 07 O.OQE+ 01 Ag-110M 4.99E+03 3.30E+04 9.97E+03 7.21E+03 1.09E+04 O.OOE+01 3.66E+06 O.OOE+01 Sn-113 4.89E+03 2.29E+03 4.67E+03 ).74E+02 9.94E+01 6.73E+01 2.30E+05 O.OOE+01 Sb-124 1.20E+04 5.91E+04 3.79E+04 5.56E+ 02 O.OOE+ 01 1.00E+ 02 2.64E+06 O.OOE+ 01 I

Te-127M 2.07E+03 2.73E+04 1.66E+04 6.89E+ 03 3.75E+04 4.86E+03 1.31E+06 O.OOE+ 01 Te-129M 2.22E+03 6.89E+04 1.41E+04 6.08E+03 3.17E+04 5,47E+03 1.68E+06 O.OOE+01 I-131 1.96E+04 1.06E+03 3.79E+04 4.43E+04 5.17E+04 1.48E+07 O.OOE+01 O.OOE+Ol I-l 32 1.26E+03 1.90E+03 1.69E+03 3.54E+03 3.94E+03 1.69E+05 O.OOE+01 O.OOE+ 01 I-133 5.59E+03 2.15E+03 1.32E+ 04 1.92E+ 04 2.24E+04 3.55E+06 O.OOE+01 O.OOE+01 I-1 35 2.77E+03 1.83E+03 3.86E+ 03 7.59E+ 03 8.46E+03 6.95E+05 O.OOE+01 O.OOE+01 Cs.134 7.44E+ 04 1.33E+ 03 3.96E+ 05 7.02E+05 1.90E+05 O.OOE+01 7.95E+04 O.OOE+01 Cs-136 5.28E+04 1.43E+03 4.82E+ 04 1.34E+05 5.63E+04 O.OOE+01 1.17E+04 O.OOE+01 Cs-137 4.54E+ 04 1.33E 4 03 5.48E+ 05 6.1 1E+ 05 1.72E+ 05 O.OOE+ 01 7.1 2E+ 04 O.OOE+ 01 Ba-140 2.89E+ 03 3.83E+04 5.59E+04 5.59E+01 1.34E+01 O.OOE+ 01 1.59E+06 O.OOE+Ol Ce-141 1.99E+03 2.15E+04 2.77E+04 1.66E+04 5.24E+03 O.OOE+ 01 5.16E+05 O.OOE+01 Ce-144 1.76E+05 1.48E+05 3.19E+06 1.21E+06 5.37E+05 O.OOE+01 9.83E+06 O.OOE+01 Hf-181 5.05E+03 1.90E+04 5.65E+04 2.66E+02 1.59E+02 2.25E+02 6.73E+05 O.OOE+01 R Values in units of mrem/yr per yCi/ms for inhalation end tritium and in units of mrem/yr per yCi/sec per m'or all others.

3-52

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 Figure 3.1 SHNPP Gaseous Waste Streams Turbine Bldg Vent Stack 3A Knt Ksnaaroa arnsoataroaaar t$yg $ $ aau yroraama Stare RlslIYJQsl(eerarr RM RAR.Roarrar AaaQarr Nrr stra Farl froarrroe arrr

%RQt.ntro Roars Oaa afoseor RAICIIYJQI E KM Karrauoa Aroairor Condensel Vaclan Pump Potisher Area 'ondenser WPB Vent Stack 5 rosrwr4$ osrreecut RM RRISIQYJllr WPB Hot tk Cold Laundry WPB Office Area Exhaust WPB Cold Laundry Oryers WPB Ccntrd Rocm Smoke Exhaust WPB Ofrace Areas WP8 General Area Exhaust WPB Chiller Room Exhaust Waste Gas Decay Tanks Waste Processfng Area Filter ed Exhaust WPB Vent Stack 5A ASSI lav4$ II I oa xulrf REta RxssrllYJIII WPB Switchgear Room Exhaust WPB HVAC Equip. Room Exhaust WPB Personnel Handeling Facility Exhaust WPB Hol rk Low ActivityLabs Exhaust WPB Lab Areas Exhaust Plant Vent Stack 1 RIWIAYJIaa Isa crlecalr Rsls I cv4$ n e RFM RessIAYJIIQA Containment Plekntry Purge RlASIAYJQI FHB Normal Exhaust North RAB Normal Exhaust FHB Normal Exhaust South RAISIAYJQIA E QÃrrcJQI RAB Emergency Exhaust Rrl'slAYJQIR FHB Normal Exhaust (Oper. Fl) South E

Rxlsl IL)Irr RAB Ventilation System FHB Normal Exhaust (Oper. Fl) North Hydrogen Purge Rxlsl ICJQIAJA RAB Smoke Purge E xrxsrrcJIaaels FHB Emergency Exhaust RAB Purge 3-53

O CD tfI Q Ct 0 III U X PLANT 0 III NO RTH CII MAGNETIC O tn

~ NORTH 25 0 C C 0 U CD I I 0 O DI 6 Ct Q 0 CD 0 Z VENT STACK D) 0 0 X Z III CII Z O I 0 pz. O C 0 QI I

UNIT III QI IU Z U Ov U

III CC 0 Tl O~

VENTSTACK D CCI g CD C

A 6 COOLING CD ITl CD z

'CI TOWER Ig 63 C M CD WAT ER SERVICE EATME ECURIT BLOC.

OO0 ZU PARKING AREA XI CD WARE EO 0 JCI CD HOUSE g QI CD CD 0

O SWITCH YARD PARKING CD QQ AREA Cl C

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 Figure 3.3 SHNPP Condenser Off-Gas System Turbine Building Vent Stack - 3A Wide Range Caa Monitor (WRCM)

RID tTV4535 t Gland Steam Condenser CVPETS" Main Condenser REM REM-tTV.3534 Regglnag Valve Atmosphere Blind Range 3-55

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Table 4.1 contains the sample point description, sampling and collection frequency, analysis type, and frequency for various exposure pathways in the vicinity of the SHNPP for the radiological monitoring program.

Figure 4.1-1 shows the exclusion boundary surrounding SHNPP.

Figures 4.1-2, 4.1-3, and 4.1-4 show the locations of the various sampling points and TLD locations. Figure 4.1-5 provides a legend for Figures 4.1-2 through 4.1-4.

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE 4.1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or Sample: Airborne Particulates and Radioiodine Sampling and Collection Frequency: Continuous operating sampler with sample collection as required by dust loading but at least once per 7 days ~

Analysis Frequency and Weekly Gross Required Analysis: I-131 (charcoal canisters)

Beta'eekly Quarterly Gamma Zsotopic'~ (Composited b location)

Sample Point ZD No. Sample Point, Description', Distance, and Direction 0.1 mi. S on SR 1134 from SR 1011 intersection.

N sector, 2.6 mi. from site.

2 1.4 mi. S on SR 1134 from SR 1011 intersection.

NNE sector, 1.4 mi. from site.

0.7 mi. N on SR 1127 from intersection with US l.

NNE sector, 3.1 mi. from site.

Pittsboro (Control Station)'NW sector from site, > 12 mi. from site Harris Lake Spillway S sector, 4.7 mi. from site 1.3 mi. N on SR 1912 from intersection of NC 42 SSW sector, 3.4 miles from site

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE 4. 1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or Sample: Direct Radiation (TLD)

Sampling and Collection Frequency: Continuous measurement with an integrated readout at least once per quarter.

Analysis Frequency and Required Analysis: Quarterly Gamma Dose Sample Point ID No. Sample Point, Description', Distance, and Direction 0.1 mi. S on SR 1134 from SR 1011 intersection.

N sector, 2.6 mi. from site.

1.4 mi. S on SR 1134 from SR 1011 intersection.

NNE sector, 1.4 mi. from site.

HE&EC ENE sector, 2.6 mi. from site.

0.7 mi. N on SR 1127 from intersection

\

with Us 1 NNE sector, 3.1 mi. from site.

Pittsboro (Control Station) ~

WNW sector from site, > 12 mi. from site Intersection of SR 1134 & SR 1135.

NE sector, 0.8.mi. from site.

Extension of SR. 1134.

E sector 0.7 mi. from site.

Dead end of road. Extension of SR 1134.

ESE sector, 0.6 mi. from site.

1 mi. S on SR 1130 from intersection of SR 1127, 1115, and 1130.

SE sector, 2.2 mi. from site.

10 SR 1130 S of intersection of SR 1127, 1115, and 1130.

SSE sector, 2.2 mi. from site.

SHNPP site.

S sector, 0.6 mi. from site 12 SHNPP site.

SSW sector, 0.9 mi. from site.

13 SHNPP site.

WSW sector 0.7 mi. from site.

SHNPP site. Access road to aux. reservoir.

W sector 1.5 mi. from site.

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 TABLE 4.1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Ex osure Pathway and/or Sample: Direct Radiation (TLD)

Sampling and Collection Frequency: Continuous measurement with an integrated readout at least once er quarter.

Analysis Frequency and Required Analysis: 'Quarterly Gamma Dose Sample Point ID No. Sample Point, Description', Distance, and Direction 15 SR 1911.

W sector, 2.0 mi. from site.

1.2 mi. E of intersection of US 1 and SR 1011.

WNW sector 1.9 mi. from site.

17 Intersection of US 1 and Aux. Res.

NW sector, 1.5 mi. from site.

18 0.2 mi. N on US 1 from Station 17.

NNW sector, 1.4 mi. from site."

19 0.6 mi. E on SR 1142 from intersection of SR 1141.

NNE sector 5.0 mi. from site.

20 US 1 at intersection SR 1149..

NE sector 4.5 mi. from site.

21 1.2 mi. W on SR 1152 from intersection SR 1153.

ENE sector, 4.8 mi. from site.

22 Formerly Ragan's Dairy on SR 1115.

E sector, 4.3 mi. from site.

23 Intersection of SR 1127 and SR 1116.

ESE sector, 4.8 mi. from site.

Sweet Springs Church on SR 1116.

SE sector 4.0 mi. from site.

25 0.2 mi. W on SR 1402 from intersection of SR 1400 SSE sector, 4.7 mi. from site Harris Lake Spillway S sector, 4.7 mi. from site 27 NC 42 8 Buckhorn United Methodist Church SW sector, 4.8 mi. from site.

28 0.6 mi. on SR 1924 from intersection of SR 1916.

SSW sector, 4.8 mi. from site.

29 Parking lot of Neste Resins Corporation on SR 1916.

WSW sector 5.7 mi. from site.

Shearon Harris Nuclear, Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 TABLE 4. 1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Ex osure Pathwa and/or Sam le: Direct Radiation (TLD)

Sampling and Collection Frequency: Continuous measurement with an integrated readout at least once per quarter.

Analysis Frequency and Required Analysis: Quarterly Gamma Dose Sample Point XD No. Sample Point, Descri tion', Distance, and Direction 30 Exit intersection of SR 1972 and US 1.

W sector, 5.6 mi. from site.

31 At intersection of SR 1908, 1909, 1910.

WNW sector, 4.7 mi. from site.

32 SR 1008.

NNW sector 6.4 mi. from site.

33 SR 1142. 1.7 mi.from intersection of SR 1141.

NNW sector, 4.5 mi. from site.

Apex (Population Center)'.

NE sector,,8.7 mi. from site.

35 Holly Springs (Population Center)

E sector, 6.9 mi. from site.

36 SR 1393't intersection of SR 1421.

E sector, 10.9 mi. from site.

37 US 401 at old CP&L office, Fuquay-Varina (Pop. Center)

ESE sector, 9.2 mi. from site.

SR 1142. 1. 5 mi. from intersection of SR 1141.

N sector, 4.5 mi. from site.

SR 1127. 0. 3 mi. S from intersection with US 1.

NNE sector, 2.5 mi. from site.

50 SR 1127 W from intersection SR 1115 and 1130.

ESE sector, 2.6 mi. from site.

53 SR 1972 N from intersection of SR 1910 and SR 1972.

NW sector, 5.8 mi. from site.

56 SR 1912 at intersection of SR 1912 and SR 1924.

WSW sector, 3.0 mi. from site.

63 SHNPP Site.

SW sector, 0. 6 mi. from site.

4-5

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 TABLE 4.1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway andlor Sample: Waterborne, Surface Water Sampling and Collection Frequency: Composite sample'ollected over a period of less than or equal to 31 days.

Analysis Frequency and Required Analysis: Monthly Gross Beta'nd Quarterly Gamma Isotopic Sample Point ID No. Sample Point, Description', Distance, and Direction 26 Harris Lake Spillway S sector, 4.7 mi. from site 38 Cape Fear Steam Electric Plant Intake Structure (Control Station) '

WSW sector, 6.2 miles from site 40 NE Harnett Metro'ater Treatment Plant Intake Building Duncan Street, Lillington, N.C.

SSE sector, -17 mi..from site.

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 TABLE 4. 1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or Sam le: Waterborne, Groundwater Sampling and Collection Frequency: Grab sample collected quarterly Analysis Frequency and Required Analysis: Each Sample Gamma Isotopic'nd Tritium Sample Point ID No. Sample Point, Description', Distance, and Direction 39 On-site deep well in the vicinity of the diabase dike SSW sector, 0.7 mi. from site 57 SHNPP Site N. bank of Emergency Service Water (ESW) intake channel, S. of Water Treatment Building.

SSW sector, 0.4 mi. from site 58 SHNPP Site N. bank of Emergency Service Water (ESW) intake channel, S. of Water Treatment Building.

WSW sector, 0.5 mi. -.from site 59 S)(NPP Site N. side of Old Construction Road.

NNE sector, 0.5 mi. from site 60 SHNPP Site W. bank of Thomas Creek N of Cooling Tower.

ESE sector, 0.5 mi. from site

Shearon Harris Nuclear Power Plant (SHNPP) December, 1997 Offsite Dose Calculation Manual (ODCM) Rev. 10 TABLE 4.1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or Sample: Waterborne. Drinking Water Sampling and Collection Frequency: Composite sample~ collected over a two-week period if I-131 analysis is performed; monthly composite otherwise.

Analysis Frequency and Required Analysis: I-131 on each composite when the dose'alculated for the consumption of the water is greater than 1 mrem per yr.

Monthly Gross Beta Monthly Gamma Tritium Isotopic'uarterly Sample Point ID No. Sample Point, Description', Distance, and Direction 38 Cape Fear Steam Electric Plant Intake Structure (Control Station)'SW sector, 6.2 miles from site 40 NE Harnett Metro Water Treatment Plant Intake Building Duncan Street, Lillington, N.C.

SSE sector, -17 mi. from site.

51 SHNPP Water Treatment Building On Site NOTE: H-3 analysis is normally performed monthly.

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 TABLE 4.1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or Sample: Waterborne, Sediment from Shoreline Sampling and Collection Frequency: Shoreline Sediment sample collected semiannually.

Analysis Frequency and Required Analysis: Each Sample Gamma Isotopic'ample Point 1D No. Sample Point, Description', Distance, and Direction 26 Harris Lake Spillway S sector, 4.6 mi. from site Shoreline of Mixing Zone of Cooling Tower Blowdown Line S sector, 3.8 miles from site.

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 TABLE 4. 1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Ex osure Pathway and/or Sample: Waterborne, Bottom Sediment Sampling and Collection Frequency: Bottom Sediment sample collected semiannually.

Analysis Frequency and Required Analysis: Each Sample Gamma Isotopic'ample Point ID No. Sam le Point, Description', Distance, and Direction 52 Harris Lake in the vicinity of the mixing zone of the cooling tower S sector, 3.8 miles from site.

4-10

Shearon Harris Nuclear Power Plant (SHNPP) December, 1997 Offsite Dose Calculation Manual (ODCM) Rev. 10 TABLE 4.1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or Sample: ingestion - Milk Sampling and Collection Frequency: Grab samples semi-monthly when animals are on pasture; monthly at other times.

Analysis Frequency and Required Analysis: Each Sample I-131 Each Sample Gamma Xsotopic'ample Point lD No. Sample Point, Description', Distance, and Direction Maple Knoll Dairy on SR 1403 SSE sector, 7.0 mi. from site is no longer in business.

Goodwin ' Dairy on SR 1134 N sector, 2.2 mi. from site is no longer in business Manco's Dairy, Pittsboro (Control Station)'NW sector from site, > -12 mi,. from site 4-11

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 TABLE 4. 1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Ex osure Pathway and/or Sample: Ingestion Fish Sampling and Collection Frequency: One sample of each of the following semiannually:

Catfish (bottom feeders)

Sunfish & Largemouth Bass (free swimmers)

Analysis Frequency and Required Analysis: Each sample - Gamma Isotopic'n edible portion for each Sample Point ID No. Sample Point, Description', Distance, and Direction Site varies within the Harris Qake.

Site varies above Buckhorn Dam on Cape Fear River (Control Station)~

4-12

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE 4. 1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or Sample: Ingestion Food Products'ampling and Collection Frequency: Samples of 3 different kinds of broadleaf vegetation monthly during the growing season Analysis Frequency and Required Analysis: Each sample - Gamma isotopic'n edible portion for each Sample Point ZD No. Sam le Point, Description', Distance, and Direction Pittsboro (Control Station)'NW sector, > 12 mi. from site SR 1189. Gunter-Morris Rd.

NNE sector, 1.7 mi. from site.

55 SR 1167, Bonsai NNW sector, 2.0 mi. from site 62 SR 1127, 0. 4 m. W of SR 1134 NE sector, 2.3 mi. from site 64 SR1127; 3/4 miles S of HEEC ENE Sector, 1.8 miles from site 4-13

Shearon Harris Nuclear Power Plant (SHNPP) December, 1997 Offsite Dose Calculation Manual (ODCM) Rev. 10 TABLE 4.1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or Sample: Aquatic Vegetation Sampling and Collection Frequency: Annually Analysis Frequency and Required Analysis: Each sample Gamma Isotopic'ample Point ID No. Sample Point, Description', Distance, and Direction 26 Harris Lake Spillway S sector, 4.7 mi. from site Shoreline of Mixing Zone of Cooling Tower Blowdown Line S sector, 3.8 miles from site.

61 Harris Lake above Holleman's Crossroads (Control Location)

E sector, 2.5 mi. from site Exposure Pathway and/or Sample: Broadleaf Vegetation Sampling and Collection Frequency: Monthly Analysis Frequency and Required Analysis: Each sample - Gamma Isotopic'ample Point ID No. Sample Point, Description', Distance, and Direction 65 Site Boundary S sector, 1.36 mi. from site 66 Site Boundary SSW sector, 1.33 miles from site

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 NOTES TO TABLE 4.1 SHNPP Radiolo ical Environmental Monitorin Pro ram Sample locations are shown on Figures 4.1-2, 4.1-3,'nd 4.1-4.

Figure 4.1-5 provides a legend for Figures 4.1-2 through 4.1-4.

Particulate samples will be analyzed for gross beta radioactivity 24 hours or more following filter change to allow for radon and thorium daughter decay. Zf gross beta activity is greater than ten times the yearly mean of the control sample station activity, a gamma isotopic analysis will be performed on the individual samples.

3. Control sample stations (or background stations) are located in areas that are unaffected by plant operations. All other sample stations that have the potential to be affected by radioactive emissions from plant operations are considered indicator stations.

Gamma isotopic analysis means the identification and quantitation of gamma-emitting radionuclides that may be attributable to effluents from plant operations.

Composite samples will be collected with equipment which is capable of collecting an aliquot at time intervals which are very short (e.g.,

every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) relative to the compositing period (e.g., monthly) .

The dose will be calculated for the maximum organ and age group, using the methodology contained. in ODCM Equation 2.2-1.

7. Based on historical meteorology (1976-1987), food product Locations 54 and 55 were added in the summer of 1988 as the off-site locations with the highest predicted D/Q values. Food product locations 43 and 46 were deleted after the 1988 growing season.

4-15

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 Figure 4.1-1 SHNPP Exclusion Boundary Plan I

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Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 Figure 4.1-2 Environmental Radiological Sampling Points

-"- - -:EMBER":"" " ': "::: ".:>" >>"; '::-~:"'""" ""4><r<.:,"@%!:":">><!.";.";:;:<gW"".'.::,',":".-<-."::"4 arns Mdear fQ>."

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Shearon Harris Nuclear power plant (SHNPP) June 1994 Of fsite Dose Calculation Manual (ODCM) Rev.

Figure 4.1-3 Environmental Radiological Sampling Points g e>>>4,e ~ .>~.. e >+A..>,', Xz, . r,,%. +P, i >A>Q<?,'S+B.... e e.,g ~~. ( P>:$ n>+

NORTH CAROLNA e~N>e~yP&n(lie:4 pgjpiQri',~~~(K~,"~ w?."<$~ >:.j~"<$~<~~gi<'e5~y.'~~i "':>'>j~<~>pQene"i;:j e

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Shearon Harris Nuclear Power Plant (SHNPP) December, 1997 Offsite Dose Calculation Manual (ODCM) Rev. 10 Figure 4.1-4 Environmental Radiological Sampling Points i;F N J

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l 4-19

Shearon Harris Nuclear Power Plant (SHNPP) December, 1997 Offsite Dose Calculation Manual (ODCM) Rev. 10 Figure 4.1-5 Environmental Radiological Sampling Points STATION REFER TO STATION REFER TO NUMBER SAMPLE TYPE FIGURE NUMBER SAMPLE TYPE FIGURE AP,AC, TL 4.1A TL 4.1-3 AP, AC, TL 4.1-4 35 TL 4.1-3 TL 4.1-4 36 TL 4.1-3 AP,AC, TL 4.1-4 37 TL 4.1-3 AP, AC, MK, FC, TL 4.1-2 38 SW, DW 4.1-2 TL '.1-4 39 GW 4.1-4 TL 4.1-4 40 SW, DW 4.1-3 TL 4.1-4 41 SS, AV

'.1-4 TL 4.1-4 42 MK 4.1-3 10 TL 4.1-4 43 DELETED 4.1-4 TL 4.1-4 44 FH 4.1A 12 TL 4.1-4 45 FH 4.1-2 13 TL 4.1-4 47 AP, AC 4.1-4 14 TL 4.1-4 48 TL 4.1-3 TL 4.1-4 49 TL 4.1-4 16 TL 4.1-4 50 TL 4.1-4 17 TL 4.1-4 51 . DW 4.1-4 18 TL 4.1-4 SD 4.1-4 19 TL 4.1-3 53 TL 4.1-2 20 TL 4.1-3, 4.1-4 54 FC 4.1-4 21 TL 4.1-3 55 FC 4.1-4 22 TL 4.1-3 56 TL 4.1-4 23 TL 4.1-3 57 GW 4.1-4 TL 4.1-3 GW 4.1-4 25 TL 4.1-3, 4.1-4 59 GW 4.1-4 26 Ap,AG,AV, ss,sw. TL 4.1-3, 4.1-4 60 GW 4.1-4 27 TL 4.1-2, 4.1-4 61 AV 4.1-3 28 TL 4.1-2, 4.1-4 62 FC 4.1-4 29 TL 4.1-2 63 TL 4.1-4 30 TL 4.1-2 64 4.1-4 31 TL 4.1-2 65 BL 4.1-4 32 TL 4.1-2 66 BL 4.1-4 33 TL 4.1-3 AC Air Cartridge FC Food Crop SD Bottom Sediment AP Air Particulate FH Fish SS Shoreline Sediment AV Aquatic Vegetation GW Groundwater SW Surface Water DW Drinkin Water MK Milk TL TLD Approximate location 4-20

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 5.0 1NTERLABORATORY COMPARISON STUDIES The objective of this program is to evaluate the total laboratory analysxs process by comparing results for an equivalent sample with those obtained by an independent laboratory or laboratories.

Environmental samples from the SHNPP environs are to be analyzed by the Harris Energy & Environmental Center (HE&EC) or by a qualified contracting laboratory. These laboratories will partxcipate at least annually in a nationally recognized interlaboratory comparison study.

The results. of the laboratories'erformances in the study will be included in the Annual Radiological Environmental Operating Report (see SHNPP ODCM Operational Requirement 4.12.3).

SHNPP E&RC will perform sample analyses for gamma-emitting radionuclides in effluent releases. The E&RC radiochemistry laboratory will participate annually in a corporate interlaboratory comparison study or an equivalent study. Radxochemical analyses of composite samples required by ODCM Operational Requirements Tables 4.11-1 and 4.11-2 will be performed by the HE&EC.

Zf the CP&L laboratory or vendor laboratory results lie at greater than three standard deviations from the "recognized value, " an evaluation will be performed to identify any recommended remedial actions to reduce anomalous errors. Complete documentation on the evaluation will be available to SHNPP and will be provided to the NRC upon request.

5-1

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 6.0 TOTAL DOSE (COMPLIANCE WITH 40 CFR 190)

ODCM Operational Requirement 3.11.4 requires that the annual dose or dose commitment to a member of the public from uranium fuel cycle sources be limited to 25 mrem for the whole body and any organ except the thyroid which is limited to 75 mrem. In addition, assessment of radiation doses to the likely most exposed member of the public from primary effluent pathways, direct radxation, and any other nearby uranium fuel cycle sources are required by ODCM Operational Requirement, Appendix F, Section F.2 to show conformance with 40 CFR 190 limits. The results of the dose assessments are submitted with the Radioactive Effluent Release Report.

If a dose assessment is made as a requirement of ODCM Operational Requirement 3.11.4, the calculation will include a complete statement of the assumptions and parameters used in calculating the dose for the report.

Dose to the Likel Most Ex osed Member of the Public 6.1.1 Effluent Pathwa s The ODCM dose equations for noble gases, iodines, particulates, and tritium provide conservative estimates because the X/Q and D/Q are historical values for the exclusion boundary distances. Because these distances are fixed and represents the closest points in the unrestricted areas to the plant, it assures that 10 CFR 50 Appendix I doses to a member of the public are unlikely to be substantially underestimated.

More realistic estimates of the actual doses from the gas and liquid effluent pathways to the likely most exposed member of the public can be obtained by using the Regulatory Guide 1.109- and WASH 1258-based NRC codes LADTAP II and GASPAR.'hese permit use of current annual average meteorology X/Q and D/Q values deri.ved from the NRC XOQDOQ (NUREG/CR-2919) Code appropriate for the specific location of the rece'ptor and the applicable exposure pathways.

6.1.2 Direct Radiation Radiation exposures of members of the public from direct radiation sources (the reactor unit and other primary system components, radwaste, radioactivity in auxiliary systems such as storage tanks, transportation of radioaotive maternal, etc.) will be determined from TLD measurements. Quarterly TLD measurements at locations within three miles of the plant center (inner ring) will be compared with the four-year, pre-operational TLD measurements using methods contained in NBS Handbook 91, "Experimental Statistics," to determine any significant contribution from direct radiation associated with plant operation.

If there is a significant direct radiation component at the TLD location in the sector containing the likely most exposed member of the public then this dose will be added to the doses from effluent pathways derived from LADTAP .II and GASPAR.

6-1

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 6.2 Dose to a Member of the Public Due to Activities Within the Site oun ar The Annual Radioactive Effluent Release Report shall include assess-ments of the radiation doses to members of the public due to activities within the site boundary. The Harris Lake is generally available for public recreational purposes year-round, and certain areas are withxn the site boundary (Figure 4.1-4).

LADTAP II and GASPAR allow the calculation of doses for special cases.

The assumptions used in the calculations, e.g., exposure time, location, and activity, will be included in the Radioactive Effluent Release Report.

6-2

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6

7. 0 LICENSEE-INITIATED CHANGES TO THE ODCM Changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained as required by Technical Specification 6.10.3.p. This documentation shall contain:

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and

2) A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or@~liability of effluent, dose, or setpoint calculations.

b) Shall become effective after review and acceptance by the PNSC and the approval of the Plant General Manager.

Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the areas of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.

7-1

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 Sheet 1 of 3 APPENDIX A Carolina Power & Light Company (CP&L) has performed the assessment of the transport and dxspersion of the effluent in the atmosphere as outlined in Pre aration of Radiolo ical Effluent Technical S ecifications or uc ear ower an s, 1978) .

e me o o ogy or xs assessment was based on guidelines presented in Regulatory Guide 1.111, Revision 1 (USNRC, 1977). The results of the assessment were to provide the relative depositions flux and relative concentrations (undepleted and depleted) based on numerical models acceptable for use in Appendix Z evaluations.

Regulatory Guide 1.111 presented three acceptable diffusion models for use in estimating deposxtion flux and concentrations. These are (1) particle-in-cell model (a variable trajectory model, based on the gradient-transport theory), (2) puff-advection model (a variable trajectory model based on the statistical approach to diffusion), and (3) the constant mean wind direction model referred to here as the straight-line trajectory Gaussian diffusion model (the most widely used model based on a statistical approach). Zt was resolved that for operational efficiency, the straight-line method described in XOQDOQ Com uter Pro ram for the Meteorolo ical Evaluation of Routine KZtuent e eases a uc ear ower a sons, I ep e er wou e use or generating the required analyses of Appendix I. To provide a more realistic accounting of the variability of wind around the plant site, standard terrain/recirculation correction factors (TCF) were used.

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 Sheet 2 of 3 APPENDIX A A twelve-year record of meteorological data was used from the on-site meteorological program at the Shearon Harris Plant. These data consisted of all collected parameters from the normal 10.0 meter tower level for the years 1976-1987. The description of the model used and the computations are presented in NUREG/CRC-2919. The following tables provide the meteorological dilution, factors (i.e., the X/Q and D/Q values) utilized to show compliance with ODCM Operational Requirements 3/4.11.2 for noble gases and radioiodines and particulates.

Tables A-1 through, A-4 Relative undepleted concentration, relative depleted concentration, and relative deposition flux estimates for round level releases for special ocations for long-term releases.

Tables A-5 through A-12 Relative undepleted concentration, relative depleted concentration, and relative deposition flux estimates for ground level releases for standard and segmented distance locations for long-term releases.

Table A-13 SHNPP on-site joint wind frequency distributions for 1976-1987.

The X/Q and D/Q values which are utilized in tne appendices are all assumed to be ground level releases'

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 Sheet 3 of 3 APPENDIX A The NRC "XOQDOQ" Program (Version 2.0) was obtained and installed on the CP&L computer system. For routine meteorological dispersion evaluations, the "XOQDOQ" Program will be run with the appropriate physical plant data, appropriate meteorological informatxon for the standard distances, and special locations of interest with a terrain/recirculation factor. The input to "XOQDOQ" for ground level releases are presented in Table A-14. The resulting computations will have the TCFs applied to produce a final atmospheric diffusion estimate for the site.

Shearon Harr's Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual tODCM) Rev. 6 TABLE A-j. THROUGH A-4 X/Q and D/Q Values for Long Term Ground Level Releases at Special Locations XOQDOQ -- SKHPP GROUND-LEVEL HISTORICAL DATA, 1976-1987 EXIT OHE -- GROUND-LEVEL HISTORICAL DATA, 1976-1987 CORRECTED USING STANDARD OPEN TERRAIN FACTORS SPECIFIC POINTS OF INTEREST X/Q

~Tb e X/Q X/Q

~T~4D/Q (sec/m3) (sec/m3) (sec/m3) (per m2)

RELEASE TYPE OF DIRECTION 0 I STAHCE NO DECAY 2.3 d DECAY 8.0 d DECAY A SITE BOUNDARY S 1. 6 2189. 6. 1E-06 .9E-06 5.2E-06 8.8E-09 A SITE BOUNDARY SSW 1.33 2140. 6.0E-06 5.8E-06 5.1E.06 8.7E-09 A SITE BOUNDARY SW 1.33 2140. 5.5E-06 5.4E-06 4.7E-06 7.0E-09 A SITE BOUNDARY WSW 1.33 2140. 4.8E-06 4.7E-06 4.1E-06 5.4E-09 A SITE BOUNDARY W 1.33 2140. 3.6E-06 3.6E-06 3.1E-06 4.16.09 A SITE BOUNDARY WHW 1.33 2140. 2.8E-06 2.7E-06 2.4E-06 3.3E-09 A SITE BOUNDARY HW 1.26 2028. 2.9E-06 2 'E-06 2.5E-06 3.96-09 A SITE BOUHDARY NNM 1.26 2028. 3.3E-06 3.2E-06 2.8E-06 5.1E-09 A SITE BOUHDARY N 1.32 2124. 3.8E-06 3.8E-06 3.3E-06 6.6E-09 A SITE BOUNDARY HHE 1.33 2140. 4.1E-06 4.0E-06 3.5E-06 7.9E.09 A SITE BOUNDARY NE 1.33 2140. 3.2E-06 3.2E-06 2.7E-06 7.36-09 A SITE BOUNDARY EHE 1.33 2140. 2.46-06 2.4E-06 2.1E-06 6.3E-09 A SITE BOUNDARY E 1.33 2140. 2.0E-06 1.9E-06 1.7E-06 3.8E-09 A SITE BOUNDARY ESE 1.33 2140. 1.8E-06 1.7E-06 1.5E-06 4.0E-09 A SITE BOUNDARY SE 1.33 2140. 2.1E-06 2.1E-06 1.8E-06 5.1E-09 A SITE BOUNDARY SSE 1.33 2140. 3.5E-06 3.4E-06 2.9E-06 6.2E-09 A NEAREST RESIDENT SSW 3.90 6275. 7.8E-07 7.3E-07 5.9E-07 6.8E-10 A HEAREST RESIDENT SW 2.80 4506. 1.36-06 1.2E-06 1.0E.06 1.2E.09 A NEAREST RESIDENT WS'M 4.30 6920. 5.4E-07 5.0E-07 4.0E-07 3.46-10 A NEAREST RESIDENT W 2.70 4345. 9.2E-07 8.8E-07 7.3E-07 7.3E-10 A NEAREST RESIDENT WHW 2.10 3380 1.1E-06 1.1E.06 9.1E-07 1.1E.09 A NEAREST RESIDENT N'W 1.80 '897.

1.4E-06 1.3E-06 1.1E-06 1.6E-09 A NEAREST RESIDENT HNW 1.50 2414. 2.2E-06 2.2E-06 1.9E.06 3.3E-09 A HEAREST RESIDENT N 2.20 3540. 1.4E-06 1.3E-06 1.1E.06 1.9E-09 A NEAREST RESIDENT NNE 1 ~ 70 2736. 2 'E-06 2.4E-06 2.0E.06 4.3E.09 A NEAREST RESIDENT HE 2.30 3701. 1.0E-06 1.0E-06 8.5E-07 1.9E.09 A NEAREST RESIDENT EHE 2.00 3219. 1.0E-06 1.0E-06 8.5E-07 2.3E-09 A NEAREST RESIDENT E 1.90 3057. 9.3E-07 9.1E-07 . 7.7E-07 1.6E-09 A NEAREST RESIDENT ESE 2.70 4345. 4.3E-07 4.1E-07 3.4E-07 7.3E-10 A NEAREST RESIDENT SE 4.70 7562. 1.9E-07 1.8E-07 1.4E-07 2.6E-10 A NEAREST RESIDENT SSE 4.40 7081. 3.6E-07 3.4E-07 2.7E-07 3.7E-10 A GARDEN SSW 3.90 6276. 7.8E-07 7.3E.07 5.9E-07 '.86-10 A GARDEN SM 2.80 4506. 1.3E-06 1.2E-06 1.0E-06 1.2E-09 A GARDEN WSW 4.30 6920. 5.4E.07 5.0E.07 4.0E-07 3.4E-IO A GARDEN W 3.00 4828. 7.6E-07 7.2E-07 6.0E-07 5.7E-10 A GARDEN WNW 2.10 3380. 1.1E-06 1.1E-06 9.1E-07 1.1E-09 A GARDEN NW 3.80 6116. 3.5E-07 3.3E-07 2.6E-07 2.8E-10 A GARDEH HHW 1.90 3058. 1.4E-06 1.4E-06 1.1E-06 1.8E-09 A GARDEN N 2.20 3540. 1.4E-06 1.3E-06 1.1E-06 1.9E-09 A GARDEN HHE 1.70 2736. 2.4E-06 2.4E-06 2.0E-06 4.3E.09 A GARDEH NE 2.30 3701. 1.0E-06 1.0E-06 8.5E-07 1.9E.09 A GARDEN E 4.70 7564. 1.86-07 1.7E-07 1.3E-07 2.0E-10 A GARDEN ESE 2.80 4506. 4.0E-07 3.9E.07 3.2E-07 6.7E-10 A GARDEN SE 4;70 7562. 1.9E-07 1.8E-07 1.4E-07 2.6E.10 A COW MILK N 2.20 3540. 1.4E-06 1.3E.06 1.1E.06 1.9E-09 A COW MILK HNE 2.92 4703. 8.5E-07 8.2E.07 6.7E.07 1.2E-09 A HEAT & POULTRY SSW 4.40 7081. 6.3E.07 5.9E-07 4.7E-07 5.2E-10 A HEAT & POULTRY SW 2.80 4506. 1.3E-06 1.2E-06 1.0E-06 1 'E-09 A HEAT & POULTRY WSW 4.30 6920. 5.4E-07 5.0E-07 4.0E-O? 3.4E-10 A HEAT & POULTRY W 3.10 4989. 7.26-07 6.8E-07 5.6E-07 5.3E-10 A HEAT & POULTRY WHW 2.50 4023. 8. 1E-07 7.7E-07 6.4E-07 7.2E-10 A HEAT & POULTRY NW 3.80 6116. 3.5E-07 3.3E-07 2.6E-07 2.8E.10 A HEAT & POULTRY NNW 1.90 3058. 1.4E-06 1.4E 06 1.1E-06 1.8E-09 A MEAT & POULTRY H 2.20 3540. 1.4E-06 1.3E-06 1.1E-06 1.9E-09 A HEAT & POULTRY NNE 1.80 2897. 2.2E-06 2.1E.06 1.8E-06 3.86.09 A MEAT & POULTRY HE 2.30 3701. 1.0E 06 1.0E.06 i 8.5E-07 1.9E-09 A MEAT & POUL'IRY ENE 2.00 3219. 1.0E 06 1.0E.06 8.5E-07 2.3E.09 A MEAT & POULTRY E 4.60 7403. 1.8E-07 1.7E.07 1.4E-07 2.1E-10 A HEAT & POULTRY ESE 4.40 7081. 1.8E.07 1.7E-07 , 1.3E-07 2.4E-10

TABLE A-5 o 0) th W W lb Undepleted, No Decay, X/Q Values for Long Term Ground Level Releases at Standard Distances (sec/m') r' rt 0 8

O X

-- 0 EXIT ONE GROUND-LEVEL HISTORICAL DATA, 1976-1987 l0 8

n~r.

NO DECAY, UHDEPLETED CORRECTED USIHG STANDARD OPEN TERRAIN FACTORS I

ANNUAL AVERAGE CHI/O (SEC/HE'IER CUBED) DISTANCE IN HILES FROH THE SITE n c 0 0 00 0 c n 0 0 4 000 V L S 1.720E-04 5.121E-OS 2.531E-OS 1.246E-OS 4.943E-06 2.814E-06 1.851E-06 1.330E-06 1.015E-06 8.077E-07 6.637E-07 0 SSM 1.571E-04 1.481E-04

4. 715E-05 4.412E-OS 2.346E-OS 2.182E-OS 1.158E-05 1.074E-OS 4.596E-06 4.263E-06 2.607E-06 2.428E-06 1.711E-06 1.597E-06 1.227E-06 1.148E-06 9.341E-07 7.424E-07 6.093E-07 r 0

g SM 8.757E-07 6.971E-OT 5.728E-07

'LISM 1.295E-04 3.842E-OS 1.893E-05 9.308E-06 3.693E-06 2.108E-06 1.389E-06 9.995E-07 7.632E-OT 6.080E-07 4.999E-07 0 M 9.839E-05 2.921E-OS 1.440E-OS 7.087E-06 2.814E-06 1.605E-06 1.05/E-06 7.608E-07 5.808E-07 4.626E-07 3.803E-OT Z 08 MH'M 7.430E.OS*- 2.216E-OS 1.098E-05 5.420E-06 2.154E-06 1.225E-06 8.053E-07 5.784E-07 4.409E-07 3.508E-07 2.881E-07 J NM 6.636E-OS 1.996E.05 9.994E-06 4.767E-06 1.982E-06 1.121E-06 7.339E-07 5.254E-07 3.995E-07 3.171E-07 2.599E-07 NNM 7. 179E-05 2.185E-OS 1. 111E-05 5.570E-06 2.234E-06 '1.254E-06 8.166E-07 5.820E-07 4.409E-OT 3.489E-07 2.853E-07 H 9.224E-OS 2.824E-OS 1.448E-OS 7.277E-06 2.919E-06 1.633E-06 1.060E-06 7.539E-07 5.701E-07 4.505E-OT 3.678E-07 HHE 9.847E-OS 3.034E-OS 1.560E-05 7.846E-06 3.145E-06 1.754E-06 1.137E-06 8.070E-07 6.094E-07 4.810E-07 3.923E-07 0 N NE 7.892E-05 2.430E-OS 1.240E-OS 6.194E-06 2.467E-06 1.376E-06 8.918E-07 6.333E-07 4.784E-07 3.777E-07 3.081E-07 0 n~

EHE 5.99SE.OS 1.845E-OS 9.388E-06 4.679E-06 1.860E-06 1.036E-06 6.711E-07 4.764E-07 3.597E-07 2.839E-OT 2.316E g lh E 4.903E-05 1.498E-OS 7 585E-06 3.779E-06 1.506E-06 8.439E-07 5.488E-07 3.908E-07 2.959E-07 2.341E-07 1.913E-07 ztg ESE 4.I58E-OS 'I.361E-OS 6.872E-06 3.417E-06 1.359E-06 7.616E-07 4.952E-07 3.527E-07 2.670E-07 2.112E-07 1.726E-07 SE 5.351E-05 1.632E-OS 8.255E-06 4.102E-06 1.629E-06 9.123E-07 5.929E-07 4.221E-07 3.195E-07 2.526E-07 2.064E-07 SSE 9. 101E. 05 2.734E-OS 1.361E-05 6.718E-06 2.666E-06 1 '09E-06 9.885E-07 7.081E-07 5.387E-07 4.278E-07 3.509E-07 ANNUAL AVERAGE CHI/O (SEC/HETER CUBED) DISTANCE IH HILES FROH THE SITE 0 0 00 0 0 00 40 0 4 0 S 5.589E-07 3.064E-07 2.075E-07 1.263E-07 8.904E-OS 6.799E-OB 5.459E-OB 4.537E-OB 3.867E-OS 3.360E-OB 2.964E-OB SS'LI 5. 126E-07 2.798E-07 1.890E-07 1.147E-07 8.066E-OB 6.149E-OB 4.930E-OB 4.093E-OB 3.485E-OB 3.026E-OB 2.667E-OB S'M 4.824E-07 2.645E-OT 1.791E-07 1.090E-07 7.686E-OB 5.869E-OS 4.712E-OS 3.916E-OB 3.338E-OB 2.900E-OB 2.558E-OB

'LIB'LI 4.213E-07 2.315E-07 1.570E-07 9.574E-OB 6.758E-OB 5.165E-OB 4.150E-OS 3.451E-OB 2.943E-OB 2.558E-OS 2.257E-OB

'LI 3.205E-07 1.760E-OT 1.193E-07 7.275E-OB 5.133E-OS 3.922E-OS 3 ~ 151E-08 2.620E-OB 2.234E-OB 1.941E-OS 1.713E-OB

'MNM 2.426E-07 1.328E-OT 8.985E-OB 5.463E-OB 3.849E-OB 2.937E-OS 2.357E-OB 1.958E-OB 1.668E-OB 1.449E.OS 1.278E-OS NM 2.185E-07 1. 188E-07 S.DOSE-OB 4.839E-OB 3.396E-OB 2.584E-OS 2.069E-OS 1.716E-OS 1.460E-OB 1.266E-OS 1. 115E-08 HNM 2.392E-07 1.290E-07 8.639E-OB 5.181E-OB 3.617E-OB 2.741E-OB 2.187E-OB 1.809E-OS 1.535E-OB 1.329E-OS 1.168E-OS N 3.081E-07 1.654E-07 1.105E-07 6.603E-OS 4.597E-OB 3.478E-OB 2.771E-OB 2.289E-OB 1.940E-OB 1.678E-OB 1.474E-OB HHE 3.283E-07 1.758E-07 1.171E-07 6.980E-OB 4.852E-OB 3.665E-OB 2.918E-OS 2.408E-OB 2.040E-OB 1.763E-OB 1.548E-OB NE 2.579E-07 1.383E-07 9.230E-OB 5.512E-OS 3.839E-OB 2.904E-OB 2.315E-OB 1.912E-OS 1.622E-OB 1.403E-OS 1.232E-OB EHE 1.938E-OT 1.039E-07 6.935E-OB 4.142E-OS 2.886E-OB 2.184E-OB 1.741E-OB 1.439E-OS 1.220E-OB 1.056E-OB 9.278E-09 E 1.604E-07 8.645E-OB 5.789E-OB '3.473E-OB 2.426E-OB 1.840E-OB 1.469E-OB 1.215E-OS 1.032E-OB 8.935E-09 7.858E-09 ESE 1.447E-07 7.804E-OB 5.228E-OS 3.139E-OS 2.193E-OB 1.664E-OB 1.329E-OB 1. 100E-08 9.341E-09 8.091E-09 7.117E-09 SE 1.730E-07 9.328E-OB 6.248E-OS 3.750E-OS 2.620E-OS 1.988E-OS 1.588E-OS 1.314E-OB 1. 116E-08 9.665E-09 8.501E-09 SSE 2.950E-07 1.608E-07 1.085E-07 6.575E-OS 4.622E-OB 3.522E-OS 2.823E-OS 2.343E-OB 1.995E-OB 1.732E-OB 1.526E-OB

TABLE A-6 0 Ph M

8 9'h IO Undepleted, No Decay, X/Q Values for Long Term Ground Level Releases at Standard Distances (sec/m') rt 0 III

-- 1976-1987 CI X EXIT ONE GROUND-LEVEL HISTORICAL DATA, 0 Na DECAY, UNDEPLETED lh O

I CHI/O (SEC/HETER CUBED) FOR EACH O tn DIRECTION SEGHENI'.824E-OS SEGHENT BOUNDARIES IN HILES FRDH THE SITE 4

I n c c n I I S 2.535E-OS 5.667E-06 1.900E-06 1.026E-06 6.676E-OT 3.185E-07 1.278E-OT 4.545E-08 3.363E-OS ft fU GI SSM 2.344E-OS 5.264E-06 1.756E-06 9.447E-07 6.129E-07 2.912E-07 1.161E-07 6.172E-OS 4.101E-08 3.029E-08 SM 2.185E-OS 4.887E-06 1.639E-06 8.855E-07 5.761E-07 2.749E-07 1. 103E-07 5.891E-08 3.923E-OS 2.903E-08 0 MSM 1.899E-OS 4.236E-06 1.425E-06 7.715E-OT 5.028E-OT 2.405E-07 9.683E-as 5.1s4E-as 3 457E-OS 2.560E-OS 0 M 1.444E-OS 3.226E-06 1.085E-06 . 5.872E-OT 3.826E-07 1.829E-OT 7.358E-OS 3.937E-OS 2.624E-08 1.943E-08 Z Q 0 8

MNM 1.099E-05 2.467E 06 8.265E-07 4.459E-OT 2.898E-07 1.381E-OT 5.528E-OS 2.948E-08 1.962E-OS 1.451E-08 c

NM NNM N

9.QTSE 06 1.103E 05 1.434E-OS 2.263E-06 2.540E-06 3.316E-06 7.538E-07 8.395E-07 1.090E-06 4.041E-07 4.462E-07 5.771E-07 2.615E-OT 2.871E-OT 3.702E-07 1.237E-OT 1.345E-07 1.727E-07 4.902E-as 5.254E-OS 6.700E-OS 2.5QSE-OS 2.753E-OS 3.494E-OS 1.71QE-08 1.813E-08 2.294E-08 1.267E-OS 1.330E-08 1.680E-OS rr NNE 1.543E-05 3.S72E-O& 1. 169E-a6 6.16QE-OT 3.949E-OT 1.836E-OT T.086E-OS 3.682E-OS 2.413E-08 1.765E-08 0 tI ~

rt NE 1.229E-OS 2.811E-06 9.176E-07 4.843E-07 3.102E-07 1.445E-OT 5.595E-OS 2.918E-08 1.917E-OS 1.404E-OS O ENE 9.309E-06 2.120E-06 6.906E-07 3.642E-07 2.331E-07 1.085E-07 4.204E-OS 2.194E-08 1.442E-OS 1.057E-08 g 0I E T.536E-06 1.T1TE-06 5.643E-OT 2.995E-07 1.925E-07 9.019E-08 3.523E-08 1.848E-OS 1.218E-08 8.946E-09 R ESE 6.833E-06 1. 551E-06 5:092E-07 2.702E-07 1.737E-OT 8.141E-08 3-183E-08 1.671E-OS 1.102E-08 8.101E-09 Ig 6.097E-07 3.233E-07 2.078E-OT 9.732E-OS 3.803E-OS 1.996E-08 1.317E-08 9.676E-09 '0 SE 8.202E-06 1.860E 06 SSE 1.360E-05 3.052E .06 1.015E-06 5.449E.07 3.530E-OT 1.674E-O7 6.657E-08 3.536E-O8 2.348E-OS '1.733E-08

TM3LE A-7 o Ul W 8 Undepleted, 2.26 Day Decay, X/Q Values for Long Term Ground Level Releases at Standard Distances (sec/m') V rt 0 0

EXIT -- GROUND-LEVEL HISTORICAL DATA, 1976-1987 O X ONE 0 2.300 DAY DECAY, UNDEPLETED N O

CORRECTED USING STANDARD OPEN TERRAIN FACtORS ANHUAL AVERAGE CHI/O (SEC/METER CUBED) DISTANCE IN MILES FROM THE SITE 0 00 nr c n S 1.712E-04 5.078E-OS 2.500E-OS 1.226E-OS 4.824E-06 2.723E-06 1. 776E-06 1 '66E-06 9. 571E-07 7.553E-07 6.153E-07 SS'M 1.564E.04 4.675E-OS 2.318E-OS 1.139E-OS 4.489E-06 2.525E-06 1.643E-06 1. 168E-06 8.820E-07 6.951E-07 5.657E-07 8 4.37/ E-05 2.154E-05 1.056E-05 4.158E-06 2.347E-06 '31E-06 1.091E-06 rt Q SW 1.474E-04 1 8.24?E-O? 6.508E-07 5.301E.07 MSM 1.289E-04 3.808E-OS 1.869E-OS 9.149E-06 3.600E-06 2.036E-06 1.330E-06 9.484E-07 7.17?E-07 5.667E-07 4.619E.07 0 W 9.794E-05 2.895E.OS 1.422E-OS 6.966E-06 2.743E-06 1;551E-06 1.013E-06 7.221E-07 5'.464E-07 4.314E-07 3.515E-07 0 WHW 7.398E-OS 2.197E-05 1.085E-OS 5.331E-06 2.102E-06 1.185E-06 7.723E-07 5 '99E-07 4.156E-07 3.278E-07 2.669E-O? Z Z O

HW 6.609E-OS 1.980E-OS 9.879E-06 4.892E-06 1.938E.06 1.087E-06 7.060E-07 5.014E-07 3.782E-07 2.978E-07 2.421E-07 NNW 7.152E-OS 2.170E-OS 1.099E-OS 5.$ 95E-06 2.189E-06 1.220E-06 7.887E-O? 5.581E-07 4.'19TE-OT 3.29/E-O? 2.6?6E-OT C H 9. 191E-05 2.805E-OS 1.434E-05 7.184E-06 2.864E-06 1.591E-06 1.026E-06 7.247E-07 5.442E-07 4 '70E-07 3;463E.O?

HNE 9. 814E-05 3.014E-OS 1.546E-OS 7.751E-06 3.088E-06 1.712E-06 1.102E-06 7.770E-O? 5.829E-07 4.570E-07 3.703E-07 HE 7.865E-05 2.414E-OS 1.228E-OS 6 '17E-06 2.422E-06 1.342E-06 8.635E-07 6.091E-07 4.569E-07 3.582E-07 2.902E.07 0 a ENE 5.9?BE-OS 1.833E-05 9.300E-06 4.621E-06 1.825E-06 1.011E-06 6.499E-07 4.582E-07 3.436E-07 2.693E-OT 2.182E-07 E 4.885E-OS 1.487E-OS 7.50?E-06 3.729E-06 1.476E-06 8.212E-07 5.301E-O? 3.748E-O? 2.81?E-07 2.212E-07 1.794E-O?

ESE 4.441E.05 1.351E-05 6.801E-06 3.372E-06 1.332E-06 '7.411E-07 4.783E-07 3.382E-07 2.541E-07 1.996E-07 1.619E-07 SE 5.331E.OS 1.621E-OS 8.172E-06 4.047E-06 1.597E-06 8.879E-07 5.728E-07 4.048E-07 3.042E-07 2.388E-07 1.937E.07 SSE 9.063E-OS 2.712E-OS 1.345E-OS 6.616E-06 2.606E-06 1.463E-06 9.505E-07 6.753E-07 5 '95E-07 4.014E-OT 3.265E-OT ANNUAL AVERAGE CHI/O (SEC/METER CUBED) DISTANCE IH MILES FROM THE SITE S 0 0 S 5.138E.07 2.700E-OT 1.755E-O? 9.871E-08 6.459E-08 4.597E-08 3./54E-08 2.697E-08 2.167E-08 'I.781E-08 1.490E-08 SS'W 4.719E-07 2.472E-O? 1.603E-07 8.990E-08 5.871E-08 4.172E-08 3.131E-08 2.442E-08 1.960E-OB 1.609E-08 1.345E 08 SW 4.426E-07 2.324E-07 1.509E-07 8.472E-08 5.532E-08 3.929E-08 2.946E-08 2.296E-08 1.841E-08 1.510E-08 1.261E-08 WSW 3.858E-OT 2.029E-O? 1.318E-07 7.405E-08 4.835E-08 3.433E-08 2.574E-08 2.005E-08 1.607E-08 1.31?E-08 1.100E-08 W 2.936E.07 1.544E-07 1.003E-O? '5.633E-08 3.678E-08 2.612E-08 1.958E-OB 1.525E-08 1.223E-OB 1.003E-OB 8.370E-09 WNM 2.228E-O? 1.169E-07 7.589E-OB 4.259E-08 2.782E-08 1.976E-08 1.482E-OB 1.155E-08 9.267E-09 7.602E-09 6-350E-09 HM 2.019E.O? 1.055E-07 6.837E-08 3.834E-08 2.505E-08 1.782E-08 1.339E-08 1.045E-08 8.394E-09 6.896E-09 5.T68E-09 NNW 2.228E-07 1.159E-OT 7.490E-08 4.194E-08 2.742E-08 1.953E-08 1-470E-08 1. 150E-08 9.252E-09 7.614E-09 6.378E.09 H 2.880E-D? 1.495E-07 9.650E-OB 5.401E-08 3.534E-08 2.520E-08 1.898E-08 1.486E-OB 1.198E-08 9.868E-09 8. 276E. 09 NHE 3.078E-OT 1.594E.OT 1.028E-O/ 5.?53E-08 3.T66E-08 2.68/E-08 2.026E-08 1.588E-08 1.281E-08 1.057E-08 8.872E-09 HE 2.413E-07 1.250E-07 8.065E-08 4.512E-08 2.953E-08 2.106E-08 1.588E-08 1.244E-08 1.003E-08 8.276E-09 6.949E.09 EHE 1.814E-07 9.395E-OB 6.060E-OB 3.391E-08 2.220E-08 1 '85E-08 1. 195E-08 9.369E-09 7.561E-09 6.239E-09 5.241E.09 E 1.493E-07 7. 761E-08 5.014E-08 2.80?E-08 1.836E-08 1.308E-08 9.847E-09 7.705E-09 6.205E-09 5.110E-09 4.284E-09 ESE 1.347E.07 7.005E-OB 4.527E-OB 2.535E-08 1.659E-08 1.183E-08 8.908E-09 6.974E-09 5.619E-09 4.630E-09 3.884E-09 SE 1.612E-07" 8.378E-OB 5.414E-OB 3.033E-08 1.985E-08 1.416E-08 1.067E-08 8.353E-09 6.732E-09 5.549E-09 4.65?E 09 SSE 2.723E-07 1.425E-07 9.243E-08 5.189E-08 3.394E-08 2.416E-08 1.816E-08 1.419E-08 1. 140E-08 9.3?BE-09 7.851E-09 C

C 8

TABLE A-8 . o 0)

&0 Undepleted, 2.26 Day Decay, X/Q Values for Long Term Ground Level Releases at Standard Distances (sec/m') 07 ill

&0 0

EXIT ONE -- GROUND-LEVEL HISTORICAL DATA, 1976-1987 U X 2.300 DAY DECAY, UNDEPLETED 0 Ol CHI/O (SEC/HETER CUBED) FOR EACH SEGHENT 8 n~

DIRECT ION SEGHENT BOUNDARIES IN NILES FROH THE SITE 4

vn 4c c 0 SSW S 2.506E-05 2.318E-05 5 '42E-06 5.151E-06 1.825E-06 1.688E-06 9.683E-07 8.926E-07 6.192E-07 5.693E-07 2.822E-07 2.585E-07 1.006E-07 9.168E-08 4.636E-OB 4.209E-08 2.711E-OS 2.455E-08 1.787E-08 1.615E-08 Vr8 S'W 2.159E-05 4.776E-06 1 '72E-06 8.345E-07 5.334E-OT 2.429E-07 8.636E-08 3.964E-OB 2.309E-08 1.516E-OS rt Q r.

1.876E-05 4.138E-06 1.366E-06 7.261E-07 q

WSW 4.647E-07 2.119E-07 7.547E-08 3.463E-OB 2.016E-08 1.323E-OS 0

'W 1.427E-OS 3.152E-06 1.040E-06 5.528E-07 3.537E-07 1.613E-07 5.741E-08 2.635E-OB 1.534E-08 1.006E-OS 0 WNW 1.087E-OS 2.412E-06 7.934E-OT 4 '05E-07 2.686E-OT 1.222E-07 4.343E-08 1.994E-OS 1.162E-08 7.632E-09 Z Z N'W 9.867E-06 2.216E-06 7.259E-07 3.827E-07 2.437E-OT 1.104E-07 O 3.911E-OS 1.798E-OB 1.051E-08 6.922E-09 NNW 1.093E-05 2.493E-06 8.116E-07 4,.250E-OT 2.694E-07 1.214E-07 4.281E-08 1.970E-08 1.156E-OS 7.641E-09 C H 1. 421E 3 '58E-06 1.056E-06 5.511E-OT 3.486E-07 1.567E-07 5.515E-08 2 '41E-08 'I.494E-08 9.903E-09 VI NNE 1.530E.05 3.513E-06 1.134E-06 5.904E-07 3.728E-07 1.672E-07 5.877E-08 2.710E 08 1.596E-08 1.060E-OS NE 1. 218E-05 2.763E-06 8.893E.07 4.628E-OT 2.923E-07 1.311E-07 4.608E-08 2.125E-08 1. 251E-08 8.305E-09 0 rt ENE 9.227E-06 2.084E-06 6.694E-07 3.480E-OT 2.197E-07 9.855E-08 3.464E-08 1.598E-08 9.418E-09 6.261E-09 aO ~

E 7.464E.06 1.685E-06 5.456E-07 2.852E-07 1.806E-07 8.134E'-08 2.866E-08 1.319E-08 7.746E-09 5.129E-09 3; th ESE 6.768E-06 1.523E-06 4.923E-07 2.574E-07 1.630E-07 7.341E-08 2.588E-08 1.193E-OS 7.010E-09 4.647E-09 SE 8.124E-06 1.826E-06 5;897E-07 3.080E-07 1.950E-07 8.781E-08 3.096E-OS 1.428E-08 8.396E-09 K

5.569E-09 IQ SSE 1.345E.05 2.989E-06 9.771E.07 5.157E-07 3.286E-07 1.491E-OT 5.292E-08 2.437E-OS 1.426E 08 9.413E-09

TABLE A-9 Q N Hl K tb O Depleted, 8.0 Day Decay, X/Q Values for Long Term Ground Level Releases at Standard Distances (sec/m') rrtN n0 Ql iD EXIT OHE -- GROUND-LEVEL HIS'IORICAL DATA, 1976-1987 CI X 8.000 DAY DECAY, DEPLETED 0 N

CORRECTED USING STANDARD OPEH TERRAIN FACI'ORS 0 I

n lo S TO ANHUAL AVERAGE CHI/O (SEC/HETER CUBED) 0 0 DISTANCE IN HILES FROH THE SITE rn c S 1.625E-04 4.665E-05 2.24?E-05 1 '85E-05 4.169E.06 2.308E-06 1.482E-06 1.042E-06 7.784E-07 6.080E-OT 4.908E-07 I r n I

SSW 1.485E-04 4.295E-05 2.083E-05 1.009E-05 3.87?E-06 .2.139E-06 1.369E-06 9.606E-07 7.168E-07 5.590E-OT 4.507E-07 Q 0 SW 1.400E-OC 4. 019E -05 1.937E-05 9.357E-06 3 '94E-06 'I.991E-06 1.278E-06 8.984E-OT 6.715E-07 5.244E-07 4.233E-07 I MSW 1.224E-OC 3.499E-05 1.681E-05 8.107E-06 3.113E-06 1.728E-06 1.1'11E-06 7.818E-07 5.849E-07 4.572E-07 3.693E-OT 0 W 9.300E-05 2.660E-05 1.279E-05 6.173E-06 2.372E-06 1.316E-06 8.458E-07 5.952E-07 4.C52E-07 3 '79E-07 2.810E-07 7.023E-05 2.018E-05 9.753E-06 0 WHW 4.721E-06 1.816E-06 1.005E-06 6.444E-OT 4.527E-07 3.382E-07 2.640E-07 2.130E-07 Z (D8 NM 6.273E-05 1.819E-05 8.876E-06 4.329E-06 1.672E.06 9.202E-07 5.878E-07 4.1'17E-07 3.068E-07 2.390E-OT 1.925E-OT NHM 6.78?E-05 1.991E-05 9.871E-06 4.85?E-06 1.886E-06 1.030E-06 6.548E-07 4.567E-07 3.391E-07 2.635E-07 2.117E-07 c H 8.721E-05 2.574E-05 1.287E.05 6.346E-06 2 '65E-06 1.342E-06 8.505E-07 5.920E-07 4 '89E-07 3.404E-07 2.732E-OT HNE 9.311E-05 2.765E-05 1.387E-05 6.844E-06 2.657E-06 1.442E-06 9.123E-OT 6.339E-07 4.694E-07 3.637E-07 2.916E-OT 7.462E-05 2.215E-OS 1. 102E-05 5.403E-06 2.084E-06 1. 131E-06 7.156E-07 4.974E-07 HE 5.671E.OS 1.682E-05 8.343E-06 4.081E-06 1.571E-06 3.683E-07 2.855E-OT 2.289E-07 0 tI rt ENE 8.519E-07 5.385E-07 3.741E-07 2.7?OE-OT 2.146E-07 4.635E-05 1.365E-05 6.739E-06 3.295E-06 1.271E-06 6.933EJOT 4.401E-07 3.067E-07 2.276E-07 1.721E-OT O~

E 4.214E-05 1.240E-05 6.105E-06 2.980E-06 1.767E-07 1.419E-07 Z M ESE 1.147E-06 6.256E-07 3.971E-07 2;76?E-07 2.054E-07 1.595E-07 1.281E-07 'z SE 5.058E-05 1.487E-05 7:335E-06 3.576E-06 1.376E-06 7.495E-07 4.755E-07 3.312E-OT 2.457E-07 1.908E-07 1.532E-O'T SSE 8.603E-05 2.491E-05 1.209E-05 5.855E-06 2.249E-06 1.238E-06 7.917E-07 5.547E-07 4.136E-07 3.223E.07 2.597E-07 tg ANHUAL AVERAGE CHI/O (SEC/HETER CUBED) DISTAHCE IN HILES FROII THE SITE 0 00 0 0 4 00 00 0 0 S 4.064E-07 2.080E-07 1.325E-07 7.280E-OS 4.700E-OS 3.317E-OB 2.477E-OB 1.924E-OB 1.538E-OB 1.25?E-OB 1.045E-OB SS'W 3.729E-07 1.901E-07 1.208E-07 6.616E-OS 4.263E-OB 3.004E-OS 2.240E-OB 1.738E-OS 1.38SE-OB 1. 134E-08 9.424E-09 S'W 3.506E-07 1.794E-07 1 ~ 143E-07 6.274E-OS 4 '49E-08 2.856E-OB 2.131E-OB 1.654E-OS 1.322E-OB 1.079E-OS 8.973E-09 MSW 3.060E-07 1.569E-07 1.001E-07 5.502E-OS 3.554E-OB 2.508E-OB 1.872E-OB 1.453E-OS 1.161E-OB 9.487E-09 7.887E-09 M 2.328E-07 1.193E-OT 7.609E-OB 4.182E-OS 2.701E-OB 1.906E-OB 1-422E-08 1.104E-OB 8.823E-09 7.207E-09 5.991E-09 WHW 1.764E.07 9. 014 E-08 5.738E.OB 3.147E-OB 2.030E-OS 1.431E. 08 1.068E-OB 8.289E-09 6.622E-09 5.409E-09 4.496E-09 NM 1.591E-07 8.086E-OS 5.129E-OS 2.801E-OB 1.803E-OB 1.269E-OB 9.462E-09 7.339E-09 5.861E-09 4.787E-09 3.979E-09 HHM 1.746E-07 B.SOBE-DS 5.560E-OB 3.019E-OB 1.936E-OB 1.360E.OB 1.013E-OB 7.848E-09 6.264E-09 5.115E-09 4.251E-09 H 2.251E.O? 1.132E-07 7.127E-OB 3.859E-OB 2.471E-OB 1.735E-OS 1-291E-08 1.000E-OB 7.981E-09 6.516E-09 5.417E-09 NNE 2.401E-07 1.204E-07 7.567E-OB 4.089E-OB 2.616E-OB 1.835E-OB 1.365E-OS 1.057E-OB 8.437E-09 6.888E-09 5.726E-09 NE 1.885E-07 9.461E-OB 5.954E-OB 3.222E-OB 2.063E-OB 1.449E-OB 1.078E-OB 8.353E-09 6.668E-09 5.445E-09 4.52?E-09 EHE 1. 41?E-07 7.109E-OB 4.473E-OS 2.421E-OB 1.551E-OB 1.089E-OS 8. 109E-09 6.285E-09 5.018E-09 4.099E-09 3.409E-09 E 1. 171E-07 5.902E-OB 3.724E-OS 2.023E-OB 1.298E-OS 9.120E-09 6.791E-09 5.264E-09 4.203E-09 3.432E-09 2.853E-09 ESE 1. 056E-07 5.327E-OB 3.363E-OS 1.827E-OS 1 ~ 173E-08 8.246E-09 6.143E-09 4.T63E-09 3.803E-09 3.107E-09 2.583E-09 SE 1.263E-07 6.369E-OB 4.020E-OS 2.184E-OS 1.402E-OB 9.856E-09 7.343E-09 5.694E-09 4.547E-09 3.714E-09 3.089E-09 SSE 2.148E-07 1.094E-07 6.946E-OB 3.801E.OS 2.449E-OS 1.726E-OB 1.288E-OS 9.994E-09 7.985E-09 6.524E-09 5.425E-09 c

W

TABLE A-10 o ol W GI Depleted, 8.0 Day Decay, X/Q Values for Iong Term Ground Level Releases at Standard Distances (sec/m') rlhn v0 0

EXIT DHE -- GROUND-LEVEL HISTORICAL DATA, 1976-1987 ax 0 P 8.00Q DAY DECAY, DEPLETED lh S

CHI/O (SEC/HETER CUBED) FOR EACH SEGMENT O 4 0 I RECT ION SEGHEHT BOUNDARIES IH HILES FROM THE SITE n c S 2.268E-05 2.09TE-Q5 4.827E-06 4.485E-06 1.526E-06 1.411E-06 7.887E-07 7.263E-07 4.943E-07 4-540E-07 2-186E-07 1.999E-QT

- 0 7.461E-OS 6.T&5E-OS 3.350E-OS 3.034E-OS

'1.935E-08 1.T4&E-OS 1.262E-OS 1.138E-OS rr c n GI rrt0 o SSM III SM 1.955E-05 4. 162E-06 1.316E-06 6.803E-07 4.264E-07 1.885E-07 6.430E-08 2.884E-08 1.664E-DS 1.084E-O&

MSM 1.698E-05 3.607E-06 1. 144E-06 5.925E-OT 3.719E.OT 1.648E-07 5.637E-08 2.532E-OS 1.462E-OS 9.524E-09 1.292E-05 2.747E-06 8.710E-07 4.51QE-07 2.830E-07 1.253E-07 4.285E-OB 1 '24E-08 1 111E-08 T.236E-09 0 MNM 9.835E-06 2.101E-06 6.63&E-07 3.426E-07 2. 145E-07 9.474E-OS 3-226E-08 1 '46E-08 8.33&E-09 5.430E-09 Z 08 III HM 8.924E 06 1.928E-06 6.060E-OT 3.109E-07 1. 939E-07 8.510E-08 2.875E-08 1 '82E-08 7.383E-09 4.806E-09 c

HNM HNE 9.874E-06 1.283E-05 1.381E-OS 2.166E-06 2.828E-06 3.048E-06 6.757E-07 8.782E-07 9.423E-07 3.439E-07 4.451E-07 4.761E-07 2.133E-OT 2.753E-07 2.939E-07 9.287E-08 1.194E-07 1.271E-07 3.102E-08 3.969E-OS 4.207E-08 1.375E-08 1.754E-OS 1.855E-OS T.&96E-09 1.006E-08 1.064E-OS 5.136E-09 6.543E-09 6.917E-09 rr NE 1.100E 05 2.39&E-06 7.392E-07 3.736E-07 2.307E-07 9.990E-OS 3.314E-08 1.464E-DS 8.404E-09 5.467E-09 On t7 ENE 8.332E-06 1.809E.06 5.563E-07 2.810E-07 1. 734E- 07 7.506E-08 2.491E-08 1 '01E-08 6.323E-09 4.115E-09 O E 6.744E-06 1.464E-06 4.542E-07 2.308E-07 1 ~ 43OE-07 6.224E-OB 2.079E-08 9.217E-09 5.296E-09 3.446E-09 3', lh ESE 6.115E 06 1.323E.06 4.099E-07 2.083E-07 1. 291E-07 5.61&E-OB 1.87&E-08 8.333E-09 4 791E-09 3. 119E-09 'z SE 7.340E-06 1.586E-06 4.90&E-07 2.492E-QT 1-544E-07 6.717E-OS 2.244E-08 9.960E-09 5.72&E-09 3.730E-09 SSE 1.217E-05 2.601E-06 8.160E-OT 4.191E.07 2.616E-OT 1.15)E-07 3.899E-08 1.743E-08 1.005E-08 6.550E-09 W

o

TABLE A-ll 0 M

&8 Deposition Values (D/Q) for Long Term Releases at Standard Distances (m ') r. g rt 0 0

EXI'I ONE -- GROUND-LEVEL HISTORICAL DATA, 1976-19S7 U X CORRECTED USING STANDARD OPEN TERRAIN FACTORS 0

M 8

~**************************** RELATIVE DEPOSITION PER UNIT AREA OI**-2) AT FIXED POINTS BY DOWNWIND SECTORS **A***A***i**********4k*****

I.

O IO

- DIRECTION DISTANCES IN HILES I

0 C 4

0 0 n 2.324E-07 7.857E-OS 4.034E-OB VV S 1.918E-OS 6.889E-09 3.417E-09 2.012E-09 1.317E-09 9.269E-10 6.869E-10 5.294E-10 A SSW 2. 171E-07 7.343E-OS 3.770E-OS 1.792E-OB 6.438E-09 3.193E-09 1.880E-09 1.231E-09 8.662E-10 6.420E-10 4.947E-10

& Ql SW 1.740E.07 5.885E-OS 3.022E-OB 1.437E-OS 5.160E-09 2.559E-09 1.507E-09 9.866E-10 6.942E-10 5.145E-10 3.965E-10 0 WSW 1.355E-07 4.582E-OS 2.353E-08 1.119E-OS 4.018E-09 1.993E-09 1. 173E-09 7.682E-10 5 406E-10 4.006E-10 3.087E-10 0 W

WNW 1.013E-07 8.296E-OS 3.427E-OS 2.805E-OS 1.759E-OS 1.440E-OS 8.365E-09 6.848E-09 3.005E-09 2.460E-09

.1.490E-09 1.220E-09 8.774E 7. 183E-10 5.745E-10 4.703E-10 4.042E-10 3.309E-10 2.996E-10 2.453E-10 2.309E-10 1.890E-10 Z Z I NW 8.468E-OS 2.864E-08 1.470E-OB 6.990E-09 2.511E-09 1.245E-09 7.332E-10 4.801E-10 3.378E-10 2.503E-10 1.929E-10 C 1.105E-07 3.736E-DB 1.918E>>OS 9'.120E-09 fig Ig HHW 3.276E-09 1.625E-09 9.566E-10 6. 264 E-10 4.408E-10 3.266E-10 2.517E-10 VV H 1.626E-07 5.499E-OS 2.823E-OS 1.342E-OS 4.821E-09 2.391E-09 1.408E-09 9.218E-10 6.487E-10 4.807E-10 3.705E-10 HNE 1.985E-07 6.713E-OB 3.447E-OS 1.639E-OS 5.886E-09 2.919E-09 1.719E-09 1. 125E-09 7.919E-10 5.869E-10 4.522E-10 Oe NE 1.815E-07 6.137E-OB 3.151E-OB 1.498E-OS 5.381E-09 2.669E-09 1.571E-09 1.029E-09 7.240E-10 5.366E-10 4.135E-10 U EHE 1.568E.07 5.302E-OB 2.723E.OS 1.294E-OB 4.649E-09 2.306E"09 1.358E-09 8.890E-10 6.255E-10 4.636E-10 3.572E-10 g Vl E 9.595E-08 3.245E-OB 1.666E-OS 7.920E-09 2.845E-09 1.411E-09 8.307E-10 5.439E-10 3.827E-10 2.837E-10 2.186E-10 'z ESE 1.003E-07 3.393E-OB 1:742E-08 8.281E-09 2.975E-09 1.475 E-09 8.686E-10 5.688E-10 4.002E-10 2.966E-10 2.286E-10 SE 1.270E-07 4.294E.OB 2.205E-08 1.048E-OB 3.765E-09 1.867E-09 1.099E-09 7.199E-10 5.066E-10 3.754E-10 2.893E-10 '0 SSE 1.548E.07 5.236E-OS 2.688E-08 1.278E-OB 4.591E-09 2.277E-09 1.341E-09 8.778E-10 6.177E-10 4.578E-10 3.528E-10 DIRECTION DISTANCES IH HILES 0 0 S 4.205E-10 1.868E-10 1. 132E-10 5.720E-11 3.462E-11 2.321E-11 1:663E-11 1.249E-11 9.711E-12 7.757E-12 6.332E-12 SSW 3.930E-10 1.746E-10 1.058E-10 5.346E-11 3.235E-11 2. 169E-11 1.554E-11 1.167E-11 9.075E-12 7.249E-12 5.917E-12 SW 3. 150E. 10 1.399E-10 8.476E-11 4.284E-11 2.593E-11 1.739E-11 1.246E-11 9.354E-12 7.273E-12 5.810E-12 4.742E-12 WSW 2.453E-10 1.090E-10 6.600E-11 3.336E-11 2.019E-11 1.354E-11 9.700E-12 7.284E-12 5.663E-12 4.524E-12 3.692E-12 W 1.834E-10 8.148E-11 4.935E-11 2.495E-11 1.510E-11 1. 012E-11 7.254E-12 5.447E-12 4.235E-12 3.383E-12 2.761E-12 WNW 1.501E-10 6.670E-11 4.040E-11 2.042E-11 1.236E-11 8.287E-12 5.938E-12 4.459E-12 3.467E-12 2.769E-12 2.261E-12 NW 1. 533E-10 6.808E-11 4.124E-11 2.085E-11 1.262E-11 8.459E-12 6.062E-12 4.552E-12 3.539E-12 2.827E-12 2.307E-12 NHW 2.000E-10 8.884E-11 5.381E-11 2.720E-11 1.646E-11 1.104E-11 7.909E-12 5.939E-12 4.618E-12 3.689E-12 3.011E-12 H 2.943E-10 1.307E-10 7.920E-11 4.003E-11 2.423E-11 1.624E-11 1.164E-11 8. 740E-12 6.796E-12 5.428E-12 4.431E-12 NHE 3.593E-10 1.596E-10 9.668E-11 4.887E-11 2.958E-11 1.983E-11 1.421E-11 1.067E-11 8.296E-12 6.627E-12 5.409E-12 HE 3.285E-10 1.459E-10 8.839E-11 4.468E-11 2.704E-11 1.813E-11 1.299E-11 9.755E-12 7.585E-12 6.059E-12 4.946E-12 EHE 2.838E-10 1.261E-10 7.637E-11 3.860E-11 2.336E-11 1.566E-11 1.122E-11 8.428E-12 6.553E-12 5.235E-12 4.273E-12 E 1. 737E-10 7.714E-11 4.673E-11 2.362E-11 1.430E-11 9.585E-12 6.868E-12 5. 157E-12 4.010E-12 3.203E-12 2.614E-12 ESE 1.816E-10 8.066E-11 4.886E-11 2.470E-II 1.495E-11 1.002E-11 7.181E-12 5.393E-12 4.193E-12 3.349E-12 2.734E-12 SE 2.298E-10 1.021E-10 6. 185E-11 3.126E-11 1.892E-11 1.269E-11 9.090E-12 6.826E-12 5.307E-12 4.240E-12 3-460E-12 SSE 2.802E-10 1. 245E-10 7.541E-11 3.812E-11 2.307E-11 1.547E-11 1.108E-11 8.323E-12 6.471E-12 5.169E-12 4. 219E-12 C

C w

TABLE A-12 0 N th 0 Deposition Values (D/Q) for Long Term Releases at Standard Distances (m ')

ct 0 O

EXIT ONE -- GROUND-LEVEL HISTORICAL DATA, 1976-1987 U X 0

(h 8

• • • • • • • • 0k***4****A*************** PER UNIT AREA (H**-2) Sy. DOWNWIND SECTORS +*********************** r-RELATIVE DEPOSITION n N DIRECTION SEGHENI'OUNDARIES IH HILES FROH THE SITE

-4 rz n c S 3.943E.OS 8.077E-09 2.109E-09 9.470E-10 5.357E-10 2.060E-10 5.960E-11 2.362E-11 1.261E-11 7.808E-12 c n I

SSW 3.6S5E.OS 7.54SE-09 1.971E-09 8.850E-10 5.007E-10 1.925E-10 5.570E-11 2.208E-11 1.179E-11 7.297E-12 SW 2.953E-OS 6.050E.09 1.579E-09 7.093E-10 4.013E-10 1.543E-10 4.464E-11 1.769E-11 9.448E-12 5.848E-12

'WSW 2.300E-08 4.7'IOE-09 1.230E-09 '.523E-10 3 ~ 124E-10 1.201E-10 3.476E-11 1.378E-11 7.357E-12 4.554E-12 0 W 1.720E-OS 3.523E-09 9.196E-10 4.130E-10 2.337E-10 8.985E-11 2.599E-11 1.030E-11 5.502E-12 3.405E-12 1.408E-OS 2.884E-09 7.528E-10 0 WHW 3.381E-10 1.913E-10 7.356E-11 2.128E-'I1 8.434E-12 4.504E-12 2.788E-12 HW 1.437E-08 2.944E-09 7.685E-10 3.451E-10 1.952E-10 7.508E-11 2.172E-11 8.609E-12 4.597E-12 2.846E-12 1.875E-OS 3.841E-09 1.003E-09 NNW HHE N 2.760E-08 3.369E-08 5.652E-09 6.901E-09 1.476E-09 1.801E-09 4.503E-10 6!627E-10 8.091E-10 2.548E-10 3.749E-10 4.577E-10 9.797E-11 1.442E-10 1.760E-10 2.834E-11

4. 171E-11.

5.092E-11 1.123E-11 1.653E-11 2.018E-11 5.998E-12 8.828E-12 1.078E-11 3.713E-12 5.464E-12 6.671E-12 rr bl HE 3.080E-OS 6.309E-09 1.647E-09 7.397E-10 4.185E-10 1.609E-10 4.655E-11 1.845E-11 9.853E-12 6.099E-12 0 rt EHE E

2.661E-08 1.628E-OS 5.451E-09 3.335E-09 1.423E-09 8.707E-10 6.391E-10 3.911E-10 3.615E-10 2.212E-10 1.390E-10 8.507E-11 4.022E-11 2.461E-11 1.594E-11 9.754E-12 8.513E-12 5.209E-12 5.269E-12 3.224E-12 n-U Q V)

ESE 1.703E-08 3.487E-09 9.104E-10 4.089E-10 2.313E-10 8.895E-11 2.573E-11 1.020E-11 5.447E-12 3.371E-12 SE 2.155E-08 4.414E-09 1;152E-09 5.176E-10 2.928E-10 1.126E-10 3.257E-11 1.291E-11 6.894E-12 4.267E-12 z'0 SSE 2.628E-08 5.383E-09 1.405E-09 6.311E-10 3 '70E-10 1.373E-10 3.972E-11 1.574E-11 8.406E-12 5.203E-12 ~Q

A-13 o N M0 Joint Wind Frequency Distribution by Pasquill Stability Classes at SHNPP N Qp rt 0 O

XOQDOQ -- SHNPP GROUND-LEVEL HISTORICAL DATA, 1976-1987 U K 0

JOINT FREQUENCY DISTRIBUTION OF WIND SPEED AND DIRECTION N tj (D

r-ATHOSPHERIC STABILITY CLASS A 0 4 S

0.000 0.000 NW I 4 0.34 1.56 3.35 0.000 0.014 0.425 0.000 0.009 0.354 0.016 0.307 0.012 0.219 0.000 0.012 0.162 0.000 0.009 0.123 0 ~ 000 0.021 0.116 0.000 0.008 0.177 0.000 0.020 0.357 0.000 0.016 0.329 0.000 0.020 0.338 0.000 0.018 0.416 0.000 0.015 0.163 0.000 0.010 0.165 0.000 0.011 0.317 0.000 0.009 0.334 0.001 0.219 4.303 rr 0

c 0 8

Ql 5.59 0.307 0.296 0.173 0.060 0.017 0.029 0.020 0.021 0.083 0.257 0.361 0.450 0.145 0.234 0.367 0.356 3.176 8.27 0.008 0.000 0.002 0.000 0.000 0.001 0.000 0.001 0.005 0.011 0.057 0.084 0.033 0.071 0.059 0.044 0.374 0 11.18 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.001 0.000 0.004 0.004 0.008 0.000 0.001 0.018 0 11.62 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 Z Q Q 8

TOTAL 0.75 0.66 0.50 0.29 0.19 0.16 0.16 0.21 0.47 0.61 0.78 0.97 0.36 0.49 0.75 0.74 8.09 rr C

'I ATHOSPHERIC STABILITY CLASS 8 N N N 0 n.

0.34 1.56 0.000 0.014 0.000 0.015 0.000 0.023 0.000 0.022 0.000 0.017 0.000 0.021 0.000 0.017 0.000 0.018 0.000 0.022 0.000 0.020 0.000 0.032 0.000 0.021 0.000 0.025 0.000 0.013 0.000 0.018 0.000 0.017 0.000 0.314 n-U Q M 3.35 0.308 0.269 0.204 0.181 0.151 0.121 0.087 0.133 0.234 0+82 0.323 0.314 0.111 0.142 0.221 0.217 3.298 5.59 0.109 0.087 0.061 0.022 0.021 0.008 0.020 0.022 0.047 0.099 0.160 0.196 0.102 0.151 0.216 0.158 1.478 z

8.27 0.006 0.000 0.001 0.000 0.000 0.000 0 ~ 000 0.001 0.004 0.005 0.031 0.026 0.032 0.053 0.028 0.019 0.205 '0 11.18 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.002 0.002 0.001 0.003 0.001 0.001 0.010 11.62 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.001 0.000 0.000 0.000 0.000 0.001 TOTAL 0.44 0.37 0.29 0.23 0.19 0.15 0.12 0.17 0.31 0.41 0.55 0.56 0.27 0.36 0.48 0.41 5.31 ATHOSPHERIC STABILITY CLASS C HAX 0.34 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 1.56 0.031 0.031 0.035 0.028 0.023 0.031 0.041 0.037 0.050 0.057 0.069 0.045 0.035 0.029 0.033 0.038 0.610 3.35 0.3I7 0.287 0.2I5 0.220 0.152 0.131 0.121 0.158 0 '61 0.332 0.375 0.369 0.175 0.192 0.246 0.261 3.874 5.59 0. 139 0.081 0.039 0.036 0.008 0.009 0.014 0.027 0.029 0.112 0.160 0.192 0.117 0.115 0.199 0 ~ 145 1.423 8.27 0.007 0.000 0.000 0.000 0.000 0.000 0.000 0.001 0.001 0.008 0.019 0.023 0.012 0.055 0.024 0.015 0. 164 11.18 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.002 0.003 0.002 0.002 0.000 0.000 0.009 11.62 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 TOI'AL 0.52 0.40 0.32 0.28 0.18 0.17 0.18 0.22 0.34 0.51 0.63 0.63 0.34 0.39 0.50 0.46 6.08 ATHOSPHERIC STABILITY CLASS D N S N 0.34 0.002 0.002 0.002 0.002 0;001 0.001 0.001 0.001 0.002 0.002 0.001 0.001 0.00'I 0.001 0.001 0.025 1.56 O.I34 0.506 0.444 0.370 0.329 0.266 0.278 0.300 0.362 '.002 0.481 0.465 0.346 0.287 0.258 0.269 0.346 5.739 3.35 1.759 2.054 1.350 0.964 0.624 0.523 0.556 0.742 1.058 1.346 1.316 1.177 0.647 0.583 0.746 1.025 16.471 C 5.59 0.556 0.491 0.330 0.114 0.066 0.076 0.100 0.190 0.220 0.376 0.550 0.462 0.260 0.402 0.523 0.558 5.271 C 8.27 0.055 0.023 0.004 0.011 0.004 0.006 0.005 0.023 0.032 0.047 0 '69 0.075 0.042 0.092 0.062 0.063 0.610 11.18 0.000 0.000 0.000 0.000 0.000 0.000 0 F 001 0.000 0.001 0.001 0.008 0.017 0.000 0.004 0 ~ 000 0.000 0.032 Ct 11.62 TOTAL 0.000 2.81 0.000 3.08 0.000 2.13 0.000 1.46 0.001 1.02 0.000 0.87 0.000 0.94 0.000 1.26 0.000 0.000 1.68 2.25 0.000 2.41 0.000 2.08 0.000 1.24 0.000 1.34 0.000 1.60 0.000 0.001 1.99 28.15 8

C r

TABLE A-13 Q M th lD Joint Wind Frequency Distribution by Pasquill Stability Classes at SHNPP fh Ql ct 0 Ib XOQDOQ -- SHHPP GROUND-LEVEL HISTORICAL DATA( 1976-1987 ax 0

JOINT FREQUENCY DISTRIBUTION OF 'HIND SPEED AND DIRECTION ill 9

ATMOSPHERIC STABILITY CLASS E O (h 0.34 N

0.005 0.007 N

0.006 0.005 0.004 0.003 0.004 0.006 0.009 0.009 0.005 0.004 0.003 0.002 U

0.003 N'

0.003 0.079 rz n c

'l.56 3.35 0.747 1.028 0.962 0.855 0.824 0.549 0.692 0.329 0.518 0.309 0.480 0.266 0.536 0.356 0.818 0.577 1 '44 1.058 1.183 1.547 0.723 1.029 0.492 0.744 0.415 0.488 0.312 0.459 0.405 0.660 0.466 10.816 0.667 10.920 r r8 c n 5.59 0.124 0.032 0.046 0.029 0.019 0.028 0.032 0.080 0.138 0.148 0.216 0.145 0.067 0.101 0.085 0.131 1.421

~

V' Ql 8.27 0.007 0.003 0.000 0.002 0.002 0.004 0.006 0.006 0.016 0.010 0.015 0.026 0.010 0.003 0.003 0.008 0.120 0 11.18 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.003 0.004 0.000 0.001 . 0.000 0-000 0.008 0 11.62 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 Z 0Z TOTAL 1.91 1.86 1.42 1.06 0.85 0.78 0.93 1.49 2.47 2.90 1.99 1.41 0.98 0.88 1.16 1.28 23.36 ATHOSPHERIC STABILITY CLASS F rr C

QP NA H N N H 0 rt 0.34 0.018 0.018 0.017 0.014 0.011 0.009 0.011 0.0,15 0.018 0.019 0.012 0.009 0.008 0.007 0 007 0 011 0 203 O 1.56 0.777 0.788 0.735 0.621 0.500 0.414 0.489 0 667 0 803 0.857 0.553 0.395 0.341 0.291 0 321 0 467 9 017 g M 3.35 0.349 0. 159 0.072 0.071 0.056 0.033 0.043 0.046 0.128 0.192 0.176 0.144 0.098 0.106 0 098 0 154 1 925 'z 5.59 0.003 0.001 0.000 0.001 0.000 0.000 0.000 0.000 0.000 0 001 0.006 0.001 0.000 0.000 0 000 0 002 0 015 tg 8.27 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.001 0.000 0.000 'Q 0.000 0.000 0.000 0.001 11.18 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0 000 0 000 0 000 11.62 0.000 0.000 0;000 0.000 0.000 0.000 0.000 0.000 0.000 0 ~ 000 0.000 0.000 0.000 0.000 0.000 0.000 0.000

- TOTAL 1.15 0.97 0.82 0.71 0.57 0.46 0.54 0.73 0.95 1.07 0.75 0.55 0.45 0.40 0.43 0.63 11.16 ATHOSPHERIC STABILITY CLASS G 0.34 0.360 0.317 0.316 0.284 0.214 0.154 0.121 0.108 0.128 0.128 0.112 0.085 0.077 0.071 0.085 0.176 2.737 1.56 1.932 1.699 1.692 1.522 1.146 0.823 0.649 0.580 0.684 0.688 0.600 0.457 0.414 0.381 0.455 0.945 14.668 3.35 0.162 0.032 0.024 0.021 0.009 0.012 0.012 0.007 0.007 0.008 0.026 0.023 0.014 0.015 0.021 0 '47 0.439 5.59 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.001 0.000 0.000 0.000 0.001 0.001 0.003 8.27 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0:000 0.000 0.000 0.000 0.000 0.000 0.000 11.18 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0 F 000 0.000 0.000 0.000 0.000 11.62 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 0.000 TOTAL 2.45 2.05 2.03 1.83 1.37 0.99 0.78 0.70 0.82 0.82 0.74 0.57 0.5'I 0.47 0.56 1.17 17.85 OVERALL HIND DIRECTION FREQUENCY N NN N N 'M NU U FREQUEHCY: 100 94 75 59 44 36 3~7 48 70 86 78 68 41 43 55 67 1000 Cl r

TABLE A-13 0 M W O Joint Wind Frequency Distribution by Pasquill Stability Classes at SHNPP rrtfll ~0 Ql O

Period of Record : 01/01/76 - 12/31/87 tI K 0

lh TOTAL HOURS CONSIDERED ARE <<<<<<<<<< 0 I

O OI I

n c Z

OVERALL MIND SPEED FREQUENCY MAX MIND SPEED (M/S): 0.335 1.565 3.353 5.588 8.270 11.176 11.623 rc n r

OI AVE MIND SPEED (M/S): 0.168 0.950 2.459 4.470 6.929 9.723 11.400 rt Ql I

WIND SPEED FREQUENCY: 3.05 41.38 41.23 12.79 1.48 0.08 0.00 0 THE CONVERSIOH FACTOR APPLIEO TO THE MIND SPEED CLASSES IS 0.447 0 Z Z eI C

DISTAHCES AHD TERRAIN HEIGHTS IN METERS AS FUNCTIONS OF DIRECTION FROM THE SITE:

I DIRECTION = S SSM SW MSW M MNM NM NNM N NHE NE EHE E ESE SE SSE DISTANCE 2189. 2140. 2140. 2140. 2140. 2140. 2028. 2028. 2124. 2140 ~ 2140. 2140. 2140. 2140. 2140. 2140. 0 rt ELEVATION 0. 0~ 0 0- 0- 0 0. 0- 0. 4. 0. 0. 0- 0~ 0- 0. U O~M g

z

'0 VEHT AHD BUILDING PARAMETERS: '0 RELEASE HEIGHT (METERS) 0.00 DIAHETER (ME'TERS) 0.00 EXIT VELOCITY (METERS) 0.00 REP. WIND HEIGHT (METERS) 10. 0 BUILDING HEIGHT (METERS) 55.0 BLDG.MIN.CRS.SEC.AREA (SQ.METERS) 2161.0 HEAT EMISSION RATE (CAL/SEC) 0.0 ALL GROUND LEVEL RELEASES.

MIND MEASURED AT 12.0 METERS.

Q Ul TABLE A-14 rb lD Vl Shearon Harris Plant Site Input Information for Continuous Ground Level Releases rt 0 A

Calculations with the NRC XOQDOQ Program 0

0 P X

lA Variable Name Format Description O Card Type Columns Value used in V XOQDOQ O lh VZ n

Card Type 1 is an array (KOPT) of options, such that 1 = DO, 0 = BYPASS. rV Gn These options remain in effect for all release points run. Thus, all release points must have the same assumptions. 0 rt Q KOPT(l ) Option to distribute calms as the first wind-speed class 0 V 0

{if calms are already distributed by direction in Card 0

Type 6, KOPT(1) = 0, and Card Type 5 is blank). If Z 0Z KOPT(1) 1, the calm values of Card Type 5 are distributed by direction in the same proportion as the C P

direction fre uenc of wind-speed class two.

KOPT(2) Option to input joint frequency distribution data as 0 0 rt ercent frequenc . U Q 0)

KOPT{3) Option to compute a sector spread for comparison with 4 centerline value in pur e calculation {Normall = 1). 'u

'o K OPT(4) Option to plot short-term X/Q values versus probab'lity 0 of occurrence {Normally = 0).

KOPT(5) Option to use cubic spline in lieu of least square function 0 for fitting intermittent release distribution (Normally =

1).

KOPT(6) Option to punch radial segment X/Q and D/Q values (Normally = 1).

KOPT{7) Option to punch output of X/Q and D/Q values of the oints of interest (Normally = 1).

KOPT{8) Option to correct X/Q and D/Q values for open terrain recirculation.

KOPT(9) Option to correct X/Q and D/Q values using site specific 0 terrain recirculation data.

TABLE A-14 0 Vl W 0 l0 Shearon Harris Plant Site Input Information for Continuous Ground Level Releases rr 0 Calculations with the NRC XOQDOQ Program 8 Card Type Columns Variable Name Format Description Value used in XOQDOQ 10 KOPT(10) Option to use desert si ma curves (Normally = 0) 0 KOPT(1 1) Option to calculate annual X/Q with 30 degree sectors for 0 North, East, South and West and 20 degree sectors for all others. (Normally = 0, and the code will use 22-1/2 de ree sectors) 1 -80 TITLM 20A4 The main title rinted at the be innin of the output. N/A 1-5 NVEL 15 The number of velocit cate ories (maximum of 14).

6-10 NSTA l5 The number of stability categories (maximum of 7) (1 always equals Pasquill stability class A, 2 = B, ..., 7 =

G).

11 - 15 NDIS l5 The numbes of distances with terrain data for each sector. The number of distances must be the same for each sector (Card T e 10) (maximum of 10).

16-20 INC l5 The increment in percent for which plotted results are 15 printed out (Normall = 15).

21 -25 NPTYPE l5 The number of titles of receptor types (cow, garden, etc.)

(Card T e 13) (maximum of ei ht) 26-30 NEXIT 15 The number of release exit oints (maximum of five).

31 -35 NCOR I5 The number of distances of site specific correction factors for recirculation (maximum of 10).

1-5 PLEV F5.0 The height (in meters, above ground level) of the 1 2.0 measured wind presented in the joint frequency data (Card Type 7). (For elevated/ground-level mixed release, use the 10-meter level winds).

0 0)

TABLE A-14 M lD Shearon Harris Plant Site Input Information for Continuous Ground Level Releases rrt w0 (h gt Calculations with the NRC XOQDOQ Program 8 ax 0

Ql Card Type Columns Variable Name Format Descnpuon Value used in 0 r.

XOQDOQ 0 lh 6-20 DECAYS(I) 3F5.0 r0 QP

'Z c

For each I: The half-life (days) used in the X/Q 101.00 I = 1,3 calculations: if DECAYS > 100, no decay will occur; if 2.30 I c n I

O DECAYS < 0, depletion factor will be used in the X/Q -8.00 rt s I

calculations; if DECAYS = 0, X/Q will not be calculated. 0 (Normally, DECAYS(1) = 101, (2) = 2.26, (3) = -8.00.) 0 Z Z 0

21 -25 PLGRAD F5.0 Plant grade elevation (feet above sea level). If PLGRAD 0.00 C

= 0.0, DIST and HT data Card Type 10 and 11 must be in meters. If PLGRAD < 0.0, DIST in miles and HT data in feet above plant grade. If PLGRAD > 0.0 above DIST O rt U

in miles and HT data in feet above sea level. A M

1 -35 CALM(l) 7F5.0 The number of hours, or percent, of calm for each BLANK zIg I = 1,NSTA stability category; if KOPT(1) = 0, insert blank card.

(Note: I ~ 1 is stabilit class A, 2 = B, ..., 7 = G).

1 -80 FREQ(K,I,J) The joint frequency distribution in hours (or percent). The K =1,16 values for 16 (K) sectors are read on each card for each I = 1,NVEL (if KOPT(1)=0) combination of wind-speed class (I) and stability class (J).

I = 2,NVEL (if KOPT(1) =1) The loop to read these value cycles first on direction J = 1,NSTA continuing in a clockwise fashion), then on wind class and finall on stabilit class.

1-5 UCOR F5.0 A correction factor applied to wind-speed classes. If 200.

UCOR < 0: no corrections will be made. If UCOR >

100: the wind-speed classes will be converted from miles/hour to meters/second.

O M TABLE A-14 W <D Ql Shearon Harris Plant Site Input Information for Continuous Ground Level Releases rt 0 Calculations with the NRC XOQDOQ Program S U X 0

Ol Card Type Columns Variable Name Format Description 0 Value used in XOQDOQ O iA 6-75 P 4 UMAX(I) 14F5.0 The maximum wind speed in each wind-speed class, in 0.75, 3.50, 7.50, 0 c either miles/hour or meters/second. (If given in i= 0 12.50, 18.50, 25.00, I I miles/hour, set UCOR ) 100.) 26.00 I

rb Card Types 8 and 9 are read in for each cor rection factor a nd distance iven, I = 1,NCOR 0 0

-80 1 VRDIST(K,I)

K=1,16 1 6F5.0 The distance in meters at which correction factors are given. These values are read in beginning with south and All Distances = 1.0 Z 8 I proceeding in a clockwise direction (maximum of 10).

r fit tg i

1-80 VRCD(K,I) 1 6F5.0 Correction factor to be applied to X/Q and D/Q values All Factors = 1.000 0 ri U

K=1,16 corresponds to distances specified in VRDIST.

Z vi Card Types 8 and 9 are repeated for the remainin distances and correction factors 'z Card Types 10 and 11 are read in for each terrain distance and hei ht iven, I = 1,NDIS 10 1 -80 DIST(K,I) 1 6F5.0 The distance in meters at which terrain heights are given. Distance = Site K=1,16 These values are read in beginning with south and Boundary Distance proceeding in a clockwise direction (maximum of ten distances).

1 -80 HT(K,I) 16F5.0 The terrain heights (in meters, above plant grade level) All Heights = 0.0 K=1,16 corresponding to the distances specified in the DIST array (Card Type 10). These values are read in the same order as the DIST array. For a given direction and distance, the terrain height should be the highest elevation between the source and that distance anywhere within the direction sector.

Card Types 10 and 11 are repeated for the remainin distances and hei hts.

Q N TABLE A-14 th K rh 0 Ol Shearon Harris Plant Site Input Information for Continuous Ground Level Releases N 0 Calculations with the NRC XOQDOQ Program 0 O X 0

Card Type Columns Variable Name Format Description Value used in (b XOQDOQ O tn VZ 12 1-25 NPOINT(I) 5I5 The number (maximum of 30) of receptor locations for a 16,15,13,2,0,11 0 i=

c n I = 1,NPTYPE particular receptor type (such as the number of cows, S

ardens. or site boundaries). rt Q Card Types 13 and 14 are read in for each receptor t pe, thus I = 1,NpTYFE 0 0

13 1-16 TITLPT(I,J) 4A4 The title (cows, gardens, etc.) of the receptor type for the Site Boundary = 16 Z 0Z receptor locations (Card Type 14) (a maximum of 16 Nearest Resident =

spaces). 15 Garden = 13 Cow Milk = 2 0 rr Goat Milk = 0 a Meat 5 Poultry = 11 Z M

'z 14 1 -80 KDIR(I,N) 8(I5,F5.0) The receptor direction and distance. KDIR is the direction See Table A-1 PTDIST(I,N) of interest, such that 1 = South, 2 = SSW......, 16 = tg N = 1,NPOINT(I) SSE, PTDIST is the distance, in meters, to the receptor location.

Card Types 13 and 14 are repeated for the remainin receptor t pes.

Card T es 15, 16 and 17 are read in for each plant release point, thus I = 1,NEXIT.

15 1 -80 TITLE(I,J) 20A4 The title for the release point whose characteristics are described on Card T es 16 and 17.

16 1-5 EXIT(I) F5.0 The vent average velocity (meters/second). (Note: if a 0 100% ground-level release is assumed, set EXIT = 0, DIAMTR = 0, and SLEV ~ 10 meters).

16 6-10 DIAMTR F5.0 The vent inside diameter (meters). 0 16 ll - 15 HSTACK(l) F5.0 The height of the vent release point (meters, plant grade 0.0 level). If release is 100% elevated, input negative of hei ht.

16- 20 HBLDG(l) F5.0 The height of the vent's building (meters, above plant 55.0 rade level).

i

+ ft

TABLE A-14 0 OI W CI VI Shearon Harris Plant Site Input information for Continuous Ground Level Releases rt 0 Calculations with the NRC XOQDOQ Program Q l3 K 0

N Card Type Columns Variable Name Format Description Value used in O XOQDOQ n~ I 21 -25 CRSEC(I) F5.0 The minimum cross-sectional area for the vent's building 2161.0 rI 0

(square meters). I ib 16 26-30 SLEV(I) F5.0 The wind height used for the vent elevated release rt Qp 10.0 (meters, above plant rade level). 0 0

16 31 -35 HEATR(I) F5.0 The vent heat emission rate (cal/sec) (Normall = 0). 0.0 Z Z III 17 RLSID(I) A1 A one letter identification for the release point. A 17 2-5 IP URGE(I) IPURGE = 1, 2 or 3 if the vent has intermittent releases.

vw Qp PQ 0 Ql The 1, 2, or 3 corresponds to DECAYS(1), DECAYS(2), 0 ct U

or DECAYS(3) (Card Type 4), respectively, whichever is A~

Q OI used as the base for intermittent release calculations (normally no decay/no deplete X/Q, such that !PURGE(l) z'o

= 1; if a vent has no intermittent releases, IPURGE = 0. ~ g 17 6-10 NPURGE(l) IS The number of intermittent releases per year for this 0 release oint.

17 11 -15 NPRGHR(l) I5 The avera e number of hours per intermittent release.

Card Types 15, 16, and 17 are repeated for the remainin release oints.

Card Types 1 - 17 may be repeated for the next case.

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 APPENDIX B Sheet 1 of 15 DOSE PARAMETERS FOR RADIOIODINES~ PARTICULATES~ AND TRITIUM This appendix contains the methodology which was used to calculate the dose parameters for radioiodines, particulates, and tritium to show compliance with ODCM Operational Requirement 3.11.2.1.b and Appendix I of 10CFR50 for gaseous effluents.

These dose parameters, P< and Rq, were calculated using the methodology outlined in NUREG 0133 along with Regulatory Guide 1.109, Revision 1. The following sections provide the specific methodology which was utilized in calculating the Pz and Rz values for the various exposure pathways (Tables 3.2-4 and 3.3-1 through 3.3-19, respectively).

B.l Calculation of P The dose parameter, Pz, contained in the radioiodine and particulates portion of Section 3.2 includes only the inhalation pathway transport parameter of the "i" radionuclide, the receptor's usage of the pathway media, and the dosimetry of the exposure. Inhalation rates and the internal dosimetry are functions of the receptor's age; however, under the exposure conditions for ODCM Operational Requirement 3.11.2.1b, the child is considered to receive the highest dose. The 4 presents the highest dose to any organ including the whole body resulting from inhalation of radionuclide "i" by a child. The following sections provide in detail the methodology which was used in calculating the P~ values for inclusion into this ODCM.

The age group considered is the child because the bases for the ODCM Operational Requirement 3.11.2.1.b is to restrict the dose to the child's thyroid via inhalation to s 1500 mrem/yr. The child's breathing rate is taken as 3700 m'/yr from Table E-5 of Regulatory Guide 1.109, Revision 1. The inhalation dose factors for the child, DFA<, are presented in Table E-9 of Regulatory Guide 1.109 in units of mrem/pCi.

The total body is considered as an organ in the selection of DFA~.

Shearon Harris Nuclear Power Plant (SHNPP) August Offsite Dose Calculation Manual (ODCM) 199'ev.

5 APPENDIX B Sheet 2 of 15 DOSE PARAMETERS FOR RADIOZODZNES~ PARTICULATES'ND TRITIUM B.l Calculation of P (continued)

The evaluation of this pathway consists of estimating the maximum dose to the most critical organ received by a child through inhalation by:

P.

ir K'BR) DFAt. (B.1-1) where:

P.

ig ~

Dose parameter for radionuclide "i" for the inhalation pathway, mrem/yr per pCi/m';

K' constant of unit conversion; 10'CilpCi; BR The breathing rate .of the'hildren's age group,/yr; The maximum organ inhalation dose factor for the children' age group for radionuclide "i," mrem/pci.

The incorporation of breathing rate of a child (3700 m'/yr) and the unit conversion factor results in the following equation:

P.

3.7 E+09 DFJh (B.1-2)

B-2

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 APPENDIX B Sheet 3 of 15 DOSE PARAMETERS FOR RADIOIODINES, PARTICULATES'ND TRITIUM B.2 Calculation of R The bases for ODCM Operational Requirement 3.11.2.3 state that conformance with the guidance in Appendix I should be shown by calculational procedures based on models and data such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated. Underestimation of the dose can be avoided by assigning a theoretical individual to the exclusion boundary in the sector with the highest X/Q and D/Q values and employing all of the likely exposure pathways, e.g., inhalation, cow milk, meat, vegetation, and ground plane.

R~ values have been calculated for the adult, teen, child, and infant age groups for the inhalation, ground plane, cow milk, goat milk, vegetable, and beef ingestion pathways. The methodology which was utilized to calculate these values is presented below.

B-3

Shearon Harris Nuclear Power Plant (SHNPP) August 1994 Offsite Dose Calculation Manual (ODCM) Rev. 5 APPENDIX B Sheet 4 of 15 DOSE PARAMETERS FOR RADIOIODINES, PARTICULATES, AND TRITIUM B.2.1 Inhalation Pathwa The dose factor from the inhalation pathway is calculated by:

R.

4 K'BR) a (DFAt) a (B.2-1) where:

K'ose R.

ig factor for each identified radionuclide "i" of the organ of interest, mrem/yr per pCi/m';

A constant of unit conversion; 10'Ci/pCi; (BR) a Breathing rate of the receptor of age group "a, "

m'/yr; (DFQ) ~ Organ inhalation 'dose factox for xadionuclide "i" fox the receptor of age group "a", mrem/pCi.

The breathing rates (BR), for the various age groups are tabulated below, as given in Table E-5 of Regulatory Guide 1.109, Revision 1.

Breathin Rate m'/ r Infant 1400 Child 3700 Teen 8000 Adult 8000 Inhalation dose factors (DF+), for the various age groups are given in Tables E-7 through E-10 of Regulatory Guide 1.109, Revision 1.

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 APPENDIX B Sheet 5 o f 15 DOSE PARAhKTERS FOR RADIOIODINES, PARTICULATES, AND TRITIUM B.2.2 Ground Plane Pathwa The ground plane pathway dose factor is calculated by:

-X t R ~

iG I.i K'K" (SF) DFG.

i (1-e ) /Xg (B.2-2) where:

R Dose factor for the ground plane pathway for each identified radionuclide "i" for the organ of interest, mrem/yr per pCi/sec per m ~;

K' constant of unit conversion; 10'Ci/pCi; Kll A constant of unit conversion; 8760 hr/year; The radiolo'gical decay constant for radionuclide "i, " sec ';

The 'exposure time, sec; 4.73 E+08 sec (15 years);

DFG; The ground plane dose conversion factor for radionuclide "i, "

mrem/hr per pCi/m~;

A tabulation of DFG, values is presented in Table E-6 of Regulatory Guide 1.109, Revision 1.

The shielding factor (dimensionless);

(A shielding factor of 0.7 is suggested in Table E-15 of Regulatory Guide 1.109, Revision 1.)

B-5

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 APPENDIX B Sheet 6 of 15 DOSE PARAMETERS FOR RADIOIODINESg PARTI CULATES J AND TRITIUM B.2.2 Ground Plane Pathwa (continued)

Factor to account for fractional deposition of radionuclide Ill II For radionuclides other than iodine, the factor I< is equal to one. For radioiodines, the value of I~ may vary. However, a value of 1.0 was used in calculating the R values in Table 3.3-2. (Reference NUREG 0133)

B-6

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 APPENDIX B Sheet 7 of 15 DOSE PARAMETERS FOR RADIOIODINES~ PARTICULATES~ AND TRITIUM B.2.3 Grass Cow or Goat Milk Pathwa The dose factor for the cow milk or goat milk pathway for each radionuclide for each organ is calculated by:

r(1-e i e) iv (

i i "Xitb r(1 e

-XE ite ) +

Biv (1-e

) ah)

(B.2-3) s E pxi where:

R, Dose factor for the cow milk or goat milk pathway, for each identified radionuclide "i " for the organ of interest,

>)c mrem/yr per pci/sec per q';

Ks A constant of unit conversion; 10'Ci/pCi; The cow's or goat's feed consumption rate, kg/day (wet weight);

U The receptor's milk consumption rate for age group "a,"

ap liters/yr; Yp The agricultural productivity by unit area of pasture feed grass, kg/m~;

The agricultural productivity by unit area of stored feed, kg/m';

B-7

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 APPENDIX B Sheet 8 of 15 DOSE PARAMETERS FOR RADZOIODZNESg PARTICULATES'ND TRITIUM B.2.3 Grass Cow or Goat Milk Pathwa (continued)

The stable element transfer coefficients, pCi/liter per pci/day; Fraction of deposited activity retained on cow's feed grass; (DFLs) a The organ ingestion dose for radionuclide "i" for the receptor in age group "a, " mrem/pci; XE I

1 The radiological decay constant for radionuclide "i,"

sec ';

The decay constant for removal of activity on leaf and plant surfaces by weathering, sec ';

5.73 E-07 sec 'l4 day half-life)';

The transport time from feed to 'cow, or goat to milk, to receptor, sec; h

The transport time from pasture, to cow or goat, to milk to receptor, sec; Period of time that sediment is exposed to gaseous e ffluents, sec; B.

iv Concentration factor for uptake of radionuclide "i" from the soil by the edible parts of crops, pCi/Kg (wet weight) per pCi/Kg (dry soil);

B-8

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 APPENDIX B Sheet 9 of 15 DOSE PARAMETERS FOR RADIOIODINES, PARTICULATES, AND TRITIUM B.2.3 Grass Cow or Goat Milk Pathwa (continued)

Effective surface density for soil, Kg (dry soil)/m~;

fP Fraction of the year that the cow or goat is on pasture; fs Fraction of the cow feed that is pasture grass while the cow is on pasture; (dimensionless).

te Period of pasture grass and crop exposure during the growing season, sec; Factor to account for fractional deposition of radionuclide "i."

'or radionuclides other than iodine, the factor I. is equal to one. For i

radioiodines, the value of I. may vary. However, a value of 1 ' was used in calculat'ing the R values in Tables 3.3-8 through 3.3-15. (Reference NUREG 0133)

Milk cattle and goats are considered to be fed from two potential sources, pasture grass and stored feeds. Following the development in Regulatory Guide 1.109, Revision 1, the value of f, was considered unity in lieu of site-specific information. The value of f was 0.667 based upon an 8-month grazing period.

Table B-1 contains the appropriate parameter values and their source in Regulatory Guide 1.109, Revision 1.

B-9

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 APPENDIX B Sheet 10 of 15 DOSE PARAMETERS FOR RADIOIODINESi PARTICULATES g AND TRITIUM B.2.3 Grass Cow or Goat Milk Pathwa (continued)

The concentration of tritium in milk is based on the airborne concentration rather than the deposition. Therefore, the Ri is based on X/Q:

R K K F+QrUap (DFLi) i 0 75 (0

~ ~ 5/H) (B.2-4) where:

R Dose factor for the cow or goat milk pathway for tritium for the organ of interest, mrem/yr per pCi/m';

Kl I I A constant of unit conversion; 10'm/kg; H Absolute humidity of the atmosphere, gm/m';

0.75 The fraction of total feed that is water; 0.5 The ratio of the specific activity of the feed grass water to the atmospheric water.

and other parameters and values are given above. A value of H = 8 grams/ meter',

was used in lieu of site-specific information.

B-10

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev.

APPENDIX B Sheet 11 of 15 DOSE PARAMETERS FOR RADIOZODINES g PARTICULATES ~ AND TRITIUM B.2.4 Grass-Cow-Meat Pathwa The integrated concentration in meat follows in a similar manner to the development for the milk pathway; there fore:

Ri B = IiK

~ QF F Uaap Ff (DFLi a e r(1-e it )

B.iv ( 1 e -Xitb)

P s Y Ph.i p yE Ei "Xitb (1 ff ) (

r(1-e -XE ite') Bi (1 e )

)

ah) (B. 2-5)

Ph,i Ys XE i

where:

R.

Dose factor for the meat ingestion pathway for radionuclide "i" for any organ of interest, mrem/yr per pCi/sec per m ';

Fr The stable element transfer coefficients, pCi/Kg per pCi/day; U

ap The receptor's meat consumption rate for age group "a, "

kg/yr; ts Transport time from slaughter to consumption, sec;

Shearon Harris Nuclear Power Plant (SHNPP) 8une 1994 Offsite Dose Calculation Manual (ODCM) Rev.

APPENDIX B Sheet 12 of 15 DOSE PARAMETERS FOR RADIOIODINESg PARTICULATES/ AND TRITIUM B.2.4 Grass-Cow-Meat Pathwa (continued) h Transport time from harvest to animal consumption, sec; te Period of pasture grass and crop exposure during the growing season, sec; Factor to account for fractional deposition of radionuclide Ili ll For radionuclides other than iodine, 1~ is equal to one. For radioiodines, the value of I~ may vary. However, a value of 1.0 was used in calculating the R values in Tables 3.3-5 through 3.3-7.

All other terms remain the same as defined in Equation B.2-3. Table B-2 contains the values which were used in calculating Rz for the meat pathway.

The concentration of tritium in meat is based on its airborne concentration rather than the deposition. Therefore, the R~ is 'based on X/Q.

R K '<QrUap (DFLl) a 0. 75 (0. 5/H) (B. 2-6) where:

R Dose factor for the meat ingestion pathway for tritium for any organ of interest, mrem/yr per pCi/m'.

All other terms are defined in Equations B.2-4 and B.2-5.

B-12

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 APPENDIX B Sheet 13 of 15 DOSE PARAMETERS FOR RADIOIODINES, PARTICULATES, AND TRITIUM B.2.5 Ve etation Pathwa The integrated concentration in vegetation consumed by man follows the expression developed in the derivation of the milk factor. Man is considered to consume two types of vegetation (fresh and stored) that differ only in the time period between harvest and consumption; therefore:

t i b)

Riv = IiK'(DFLi)a (UafLe L '

~itL r(1-e Ei e) +

iv (

p),.

) +

Y i

S Uaf e ih

),.t (

.(1-.

v Ei )

+'X -A.itb (B. 2-7) where:

R.

iy Dose factor for vegetable pathway for radionuclide "i" for the organ of interest, mrem/yr per pCi/sec per m';

A constant of unit conversion; 10'pCi/pCi; B-13

Shearon Harris Nuclear Power Plant (SHNPP) June Offsite Dose Calculation Manual (ODCM) 199'ev.

APPEND1X B Sheet 14 of 15 DOSE PARAMETERS FOR RAD1OZOD1NES, PARTICULATES, AND TRITIUM B.2.5 Ve etation Pathwa (continued)

U" a

The consumption rate of fresh leafy vegetation by the receptor in age group "a, " kg/yr; Us a

The consumption rate of stored vegetation by the receptor in age group "a," kg/yr; The fraction of the annual intake of fresh leafy vegetation grown locally; 1.0 The fraction of annual intake of stored vegetation grown locally;

0. 76 The average time between harvest of leafy vegetation and its consumption, sec; The average time between harvest of stored vegetation and its consumption, sec; Yy The vegetation a real density, kg/m';

Period of leafy vegetable exposure during growing season, sec; Factor to account for fractional deposition of radionuclide II i II All other factors as defined before.

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev.

APPENDIX B Sheet 15 o f 15 DOSE PARAMETERS FOR RADIOIODINESg PARTICULATES J AND TRITIUM B.2.5 Ve etation Pathwa (continued)

For radionuclides other than iodine, the factor I; is equal to one. For radioiodines, the value of I~ may vary. However, a value of 1 ~ 0 was used in Tables 3.3-2 through 3.3-4.

Table B-3 presents the appropriate parameter values and their source in Regulatory Guide 1.109, Revision 1.

In lieu of site-specific data default values for f and f~, 1.0 and 0 '6, respectively, wexe used in the calculations on Rz. These values were obtained from Table E-15 of Regulatory Guide 1.109, Revision 1.

The concentration of tritium in vegetation is based on the aixborne concentration rather than the deposition. Therefore, the R~ is based on X/Q:

RT V

= K '

fL + U f; (DFL ) 0. 75 (0. 5/H) (B.2.8) where:

R Dose factor for the vegetable pathway for tritium for any organ of interest, mrem/yr per pci/m .

All other terms remain the same as those in Equations B.2-4 and B.2-7.

B-15

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 TABLE B-1 Parameters For Cow and Goat Milk Pathways Reference Parameter Value (Re . Guide 1.109, Rev. 1)

Q (kg/day) 50 (cow) Table E-3 6 (goat Table E-3 Y (kg/M') 0.7 Table E-15 p

t (seconds) 1.73 E+05 (2 days) Table E-15 r 1.0 (radioiodines) Table E-15 0.2 (particulates) Table E-15 (DFL.)

ia (mrem/pCi) Each radionuclide Table E-11 to E-14 F

,m (pCi/liter per pCi/day) Each stable element Table E-1 (cow)

Table E-2 (goat)

T (seconds) 4.75 E+08 (15 yr) Table E-15 Y

s (kr/m') 2.0 Table E-15 t (seconds) 7.78 E+06 (90 days) Table E-15 U

ap (liters/yr) 330 infant Table E-S 330 child Table E-5 400 teen Table E-5 310 adult Table E-5 te (seconds) 2.59 E+06 (pasture) Table E-15 5.18 E+06 (stored feed)

B.

iv (pCi/kg [wet weight] Each stable element Table E-1 per pCi/kg [dry soil) )

P (kg dry soil/m') 240 Table E-15 B-16

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 TABLE B-2 Parameters For The Meat Pathway Reference Parameter (Re . Guide 1.109, Rev. 1) 1.0 (radioiodines) Table E-15 0.2 (particulates) Table E-15 F (pCi/ke per (pCi/Day) Each stable element Table E-1 U

ap (kg/yr) 0 infant Table E-S 41 child Table E-5 65 teen Table E-5 110 adult Table E-5 i

(DFL.) a (mrem/pCi) Each radionuclide Tables E-11 to E-14 Y (kg/m') 0.7 Table E-15 P

Y s

(kr/m') 2.0 Table E-15 Tb (seconds) 4.73 yr) Table E-15 T (seconds) s soil/m'alue (seconds) 1.73 7.78 E+08 (15 E+06 (20 days)

E+06 (90 days)

Table E-15 Table E-15 t (seconds)' 2.59 E+06 (pasture) Table E-15 5.18 E+06 (stored feed)

Q (kg/day) 50 Table E-3 B.

iv (pci/kg (wet weight]

per pCi/kg [dry soil] )

Each stable element Table E-1 P (kg dry 240 Table E-15 B-17

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 TABLE B-3 Parameters for The Vegetable Pathway Reference Parameter Value (Re . Guide 1.109, Rev. 1) r (dimensionless) 1.0 (radioiodines) Table E-1 0.2 (particulates) Table E-1 i

(DFL.) a (mrem/pci) Each radionuclide Tables E-11 to E-14 Q (kg/day) 50 (cow) Table 6 (goat) Table E-3 U"

a (kg/yr) Infant 0 Table Child 26 Table E-5 Teen 42 Table E-5 Adult 64 Table U'kr/hr) a Xnfant 0 Table Child 520 Table Teen 630 Table Adult 520 Table T (seconds) 8.6 E+04 (1 day) Table E-15 t (secondsf 5.18 E+06 (60 day) Table E-15 Y (kg/m~) 2.0 Table E-15 V

te (seconds) 5.18 E+06 (60 day) Table E-15 t (seconds) 4.73 E+08 (15 yr) Table E-15 P (kg dry soil/m~ 240 Table E-15 B.

iV (pCi/kg [wet weight) Each stable element Table E-1 per pCi/kg (dry soil] )

B-18

Shearon Harris Nuclear Power Plant (SHNPP) June 1994 Offsite Dose Calculation Manual (ODCM) Rev. 4 Sheet 1 of 1 APPENDIX C RADIOACTIVE LIQUID AND GASEOUS EFFLUENT MONZTORZNG INSTRUMENTATZON NUMBERS Monitor Li uid Effluent Monitorin Instruments Identification A. Treated Laundry and Hot Shower Tank REM-1WL-3540 B. Waste Monitor Tank REM-21WL-3541 Waste Evaporator Condensate Tank REM-21WL-3541 D. Secondary Waste Sample Tank REM-21WS-3542 E. NSW Returns to Circulating Water System from Waste Processing Building ~ . REM-1SW-3500A from Reactor Auxiliary Building REM-1SW-3500B Outdoor Tank Area Drain Transfer Pump t

Monitor REM-1MD-3530 G. Turbine Building Floor Drains Effluent REM-1MD-3528 II. Gaseous Effluent Monitorin, Instruments A. Plant Vent Stack 1 REM-1AV-3509-SA

• RM-21AV-3509-1SA B. Turbine Building Vent Stack 3A

• RM-1TV-3536-1 Waste Processing Building Vent Stack 5 REM-1WV-3546

• RM-1WV-3546-1 Waste Processing Building Vent Stack 5A REM-1WV-3547

• RM-1WV"3547-1 Wide-Range Gas Monitor (WRGM)

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 APPENDIX D PROGRAMMATIC CONTROLS The surveillance and operational requirements pertaining to the ODCM Operational Requirements are detailed in Sections:

D.1 Instrumentation D.2 Radioactive Effluents D.3 Radiological Environmental Monitoring

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 D.l INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.10 Radioactive Li uid Effluent Monitorin Instrumentation OPERATIONAL REQUIREMENT 3.3.3.10 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their Alarm/Trip setpoints set to ensure that the limits of Operational Requirement 3.11.1.1 are not exceeded. The Alarm/Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY: At all times.

ACT1ON:

With a radioactive liquid effluent monitoring instrumentation channel Alarm/Trip Setpoint less conservative than required by the above Operational Requirement, immediately (1) suspend the release of radioactive liquid effluents monitored by the affected channel or (2) declare the channel inoperable and take ACTION as directed by b.

below.

b. 'ith less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12. Exert best effort to return the instrument to OPERABLE status within 30 days. and, if unsuccessful, explain in the next Annual Radioactive Effluent Release Report pursuant to ODCM, Appendix F, Section F.2 why this inoperability was not corrected in a timely manner.

SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and DIGITAL CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-8.

Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.

D-2

Shearon Harris Nuclear Power Plant (SHNPP) August 1996 Offsite Dose Calculation Manual (ODCM) Rev. 9 B 3 3-DIO CT V UDFU0 0G T ION MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste Effluent Lines Treated Laundry and Hot Shower 35 Tanks Discharge Monitor

2) Waste Monitor Tanks and Waste 35 Evaporator Condensate Tanks Discharge Monitor

3) Secondary Waste Sample Tank 35, 36*

Discharge Monitor

b. Turbine Building Floor Drains Effluent 36 Line Radioactivity Monitor Providing Alarm and Automatic Stop Signal to Discharge Pump

a. Outdoor Tank Area Drain Transfer Pump 37 Monitor

3. Radioactivity Monitors Providing Alarm But Not Providing Automatic Termination of Release

a. Normal Service Water System Return 39 From Waste Processing Building to the Circulating Water System

b. Normal Service Water System Return 39 From the Reactor Auxiliary Building to the Circulating Water system Flow Rate Measurement Devices a ~ Liquid Radwaste Effluent Lines

1) Treated Laundry and Hot Shower 38 Tanks Discharge

2) Waste Monitor Tanks and Waste 38 Evaporator Condensate Tanks Discharge

3) Secondary Waste Sample Tank 38 Cooling Tower Blowdown 38 When the Secondary Waste System is in the continuous release mode And releases are occurring, Action 36 shall be taken when the monitor is inoperable. In the batch release mode, Action 35 is applicable.

D-3

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE 3.3-12 (Continued)

ACTION STATEMENTS ACTION 35 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that prior to initiating a release:

a ~ At least two independent samples are analyzed in accordance with Operational Requirement 4.11.1.1.1, and At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 36 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed for radioactivity at a lower limit of detection of no more than 10 'icroCurie/ml:

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the specific activity of the secondary coolant is greater than 0.01 microCurie/gram DOSE EQUIVALENT I-131 or,

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the specific activity of the secondary coolant is less than or equal to 0.01 microCurie/gram DOSE EQUIVALENT Z-131.

ACTION 37 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, grab samples are collected and analyzed for radioactivity at a lower limit of detection of no more than 10 'icroCurie/ml.

ACTION 38 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump performance curves generated in place may be used to estimate flow.

ACTION 39 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the weekly Cooling Tower Blowdown weir surveillance is performed as required by Operational Requirement 4.11.1.1.1. Otherwise, follow the ACTION specified in ACTION 37 above.

D-4

Shearon Harris Nuclear Power Plant (SHNPP) November 1995 Offsite Dose Calculation Manual (ODCM) ,Rev. 7 TABLE 4.3-8 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS DIGITAL CHANNEL INSTRUMENT CHANNEL SOURCE CHANNEL OPERATIONAL CHECK CHECK CALIBRATION TEST

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste Effluent Lines

1) Treated Laundry and Hot R(3)

Shower Tanks Discharge Monitor

2) Waste Monitor Tanks and R(3) Q(1)

Waste Evaporator Condensate Tanks Discharge Monitor

3) Secondary Waste Sample Tank P, M(5) R(3) Q (1)

Discharge Monitor

b. Turbine Building Floor Drains R(3)

Effluent Line

2. Radioactivity Monitor Providing Alarm and Automatic Stop Signal to Discharge Pump

a. Outdoor Tank Area Drain R(3) Q (1)

Transfer Pump Monitor

3. Radioactivity Monitors Providing Alarm But Not Providing Automatic Termination of Release

a. Normal Sezvice Water System . R(3) Q(2)

Return From Waste Processing Building to the Circulating Water System

b. Normal Service Water System M R(3) Q(2)

Return From the Reactor Auxiliary Building to the Circulating Water System Flow Rate Measurement Devices

a. Liquid Radwaste Effluent Lines

1) Treated Laundry and Hot D (4) N.A. N.A.

Shower Tanks Discharge

2) Waste Monitor Tanks and D (4) N.A.

Waste Evaporator Condensate Tanks Discharge

3) Secondary Waste Sample Tank D(4) N.A. N.A.

Pump Monitor

b. Cooling Tower Blowdown D(4)

D-5

Shearon Harris Nuclear Power Plant (SHNPP) December, 1997 Offsite Dose Calculation Manual (ODCM) Rev. 10 TABLE 4.3-8 (Continued)

TABLE NOTATIONS (1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate automatic isolation of this pathway (or, for the Outdoor Tank Area Drains Monitor, automatic stop signal to the discharge pump) and control room alarm annunciation* occur if any of the following conditions exists (liquid activity channel only):

Instrument indicates measured levels above the Alarm/Trip Setpoint, Circuit failure (monitor loss of communications (alarm only),

detector loss of counts (Alarm only) and monitor loss of power),

Detector check source test failure (alarm only),

Detector channel out of service (alarm only),

Monitor loss of sample flow (alarm only). (Not applicable for Turbine Building Drain Rad Monitor)

(2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation* occurs (liquid activity channel only):

if any of the following conditions exists

a. Instrument indicates measured levels above the Alarm Setpoint, Circuit failure (monitor loss of'ommunications , detector loss of counts, and monitor loss of power),

c. Detector check source test failure,

d. Detector channel out of service,

e. Monitor loss of sample flow.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous, periodic, or batch releases are made.

(5) When the Secondary Waste System is being used in the batch release mode, the source check shall be prior to release. When the system is being used in the continuous release mode, the source check shall be monthly.

• Control Room Alarm Annunciation shall consist of a change in state of the tested channel on the RM-11 terminal (i.e., a change in color).

D-6

Shearon Hazris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3/n.3.3 MONITORING INSTRUMENTATION 3/4.3.3.11 Radioactive Gaseous Effluent Monitorin Instrumentation OPERATIONAL REQUIREMENT 3.3.3.11 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their Alarm/Trip Setpoints set to ensure that the limits of Operational Requirements 3.11.2.1 and 3.11.2.5 are not exceeded. The Alarm/Trip Setpoints of these channels meeting Operational Requirement 3.11.2.1 shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.

APPLICABILITY: As shown in Table 3.3-13 ACTXON:

With a radioactive gaseous effluent monitoring instrumentation channel Alarm/Trip Setpoint less conservative than required by the above Operational Requirement, immediately (1) suspend the release of radioactive gaseous effluents monitored by the affected channel or (2) declare the channel inoperable and take ACTION as directed by b.

below.

b. With the number of OPERABLE radioactive gaseous effluent monitoring instrumentation channels less than the Minimum Channels OPERABLE, take the ACTION shown in Table 3.3-13. Exert best efforts to return the instrument to OPERABLE status within 30 days. Zf unsuccessful, explain in the next Annual Radioactive Effluent Release Report pursuant to ODCM, Appendix F, Section 'F.2 why this inoperability was not corrected in a timely manner.

SURVEILLANCE REQUIREMENTS 4.3.3.11 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and a DXGITAL CHANNEL OPERATXONAL TEST or an ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-9.

Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.

D-7

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE 3.3-13 RADXOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATXON MIN. CHAPELS OPEWBLE APPLICABILITY ACTION INSTRUMENT GASEOUS WASTE PROCESSING SYSTEM HYDROGEN AND OXYGEN ANALYZERS Specification is not used in ODCM

2. TURBINE BUILDING VENT STACK

a. Noble Gas Activity Monitor

b. Iodine Sampler 49 C. Particulate Sampler

d. Flow Rate Monitor

e. Sampler Flow Rate Monitor

3. PLANT VENT STACK a ~ Noble Gas Activity Monitor Zodine Sampler Particulate Sampler Flow Rate Monitor.

l Sampler Flow Rate Monitor 46 WASTE PROCESSING BUILDING VENT STACK 5 a.l Noble Gas Activity Monitor (PIG) 45, 51 a.2 Noble Gas Activity Monitor (WRGM) MODES I, 2, 3 52 Zodine Sampler c Particulate Sampler Flow Rate Monitor

e. Sampler Flow Rate Monitor

5. WASTE PROCESSING BUILDING STACK 5A Noble Gas Activity Monitor Iodine Sampler Particulate Sampler Flow Rate Monitor Sampler Flow Rate Monitor TABLE NOTATXONS

• At all times.

D-8

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE 3.3-13 (Continued)

ACTION STATEMENTS ACTZON 45 With the number channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the waste gas decay tank(s) may be released to the environment provided that prior to initiating the release:

a ~ At least two independent samples of the tank's contents are analyzed, and At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve lineup.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTXON 46 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

ACTION 47 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTXON 48 Not Used in the ODCM ACTXON 49 With the number of chan'nels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples are continuously collected with auxiliary sampling equipment as required in Table 4.11-2.

ACTXON 50 - Not used in the ODCM.

ACTXON 51 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement for both the PZG and WRGM, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 52 With the number of OPERABLE accident monitoring instrumentation channels for the radiation monitor(s) less than the Minimu'm Chanels OPERABLE requirements of Technical Specification Table 3.3.10, initiate the preplanned alternate method of monitoring the appropriate parqameter(s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and either restore the inoperable channel(s) to OPERABLE status within 14 days or prepare and submit a Special Report to the Commission, pursuant to Technical Specification 6.9.2, within the next 14 days that provides actions taken, cause of the inoperability, and the plans and schedule for restoring the channel(s) to OPE~LE status.

D-9

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE 4.3-9 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS MODES FOR WHICH DIGITAL CKPÃHEL SURVEILLANCE CHANNEL SOURCE CHANNEL CPEEEEICKEE ~IS IIE IIIED INSTRUMENT CIIECK CHECK CALIBRATION TEST

1. GASEOUS WASTE PRCCESSING SYSTEM HYDROGEN AND OXYGEN ANALYZERS Not Used ln the ODCM.

2~ TURBIHE BUILDING VENT STACK

a. Noble Gas Activity R(3) Q(2)

b. Iodine Sampler N.A. H.A. N.A.

c. Particulate Sampler

d. Flow Rate Monitor D N.A.

.e. Sampler Flow Rate Monitor D N.A.

3. PLANT VENT STACK

a. Noble Gas Activity Monitor D M R(3) Q(2)

b. Iodine Sampler N>A. N.A. N.A.

c. Particulate Sampler NEA. H.A.

d. Flow Rate Monitor D N.A.

e. Sampler Flow Rate Monitor D N.A.

4. WASTE PRCCESSING BUILDING VENT STACK 5 a.l Noble Gas Activity Monitor D R(3) Q(1)

(PIG) a.2 Noble Gas Activity Monitor M R(3) Q(2)

(WRGH)

b. Iodine Sampler

c. Particulate Sampler N.A. N.A.

d. Flow Rate Monitor

e. Sampler Flow Rate Monitor D WASTE PROCESSING BUILDIHG VENT STACK SA

a. Noble Gas Activity Monitor R(3) Q(2)

b. Iodine Sampler N.A.

c. Particulate Sampler N.A. N.A.

d. I'low Rate Monitor

e. Sampler Flow Rate Monitor At all times.

D-10

Shearon Harris Nuclear Power Plant (SHNPP) August 1996 Offsite Dose Calculation Manual (ODCM) Rev. 9 (1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room following conditions exists (gas activity annunciation* occur and gas if any of the effluent channels only):

a ~ Instrument indicates measured levels above the Alarm/Trip Setpoint,

b. Circuit failure (monitor loss of communications (alarm only),

detector loss of counts (alarm only) and monitor loss of power),

c ~ Detector check source test failure (gas activity channel only),

(alarm only),

d. Detector channel out of service (alarm only),

e. Monitor loss of sample flow (alarm only).

(2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation* occurs if any of the following conditions exists (gas activity and gas effluent channels only):

a. Instrument indicates measured levels above the Alarm Setpoint,

b. Circuit failure (monitor loss of communications (alarm only),

detector loss of counts, and monitor loss of power),

c. 'etector check source test failure (gas activity channel only),

d. Detector channel out of service,

e. Monitor loss of sample flow.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with,NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(4) Not used in the ODCM.

(5) Not used in the ODCM.

• Control Room Alarm Annunciation shall consist of a "change in state of the tested channel on the RM-11 terminal (i.e., a change in color).

Shearon Harris Nuclear Power Plant (SHNPP) August 1gg5 Offsite Dose Calculation Manual (ODCM) Rev. 6 D.2 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 Concentration OPERATIONAL REQUIREMENT 3.11.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Technical Specification Figure 5.1-3) shall be limited to 10 times the concentrations specified in 10 CFR Part 20.1001 20.2401, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10 'icroCurie/ml total activity.

APPLICABILITY: At all times.

ACTION:

With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, immediately restore the concentration to within the above limits.

SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall5e sampled and analyzed according to the sampling and analysis program of Table 4.11-1.

4.11.1.1.2 The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point, of release are maintained within the limits of Operational Requirement 3. 11. 1. 1.

Each Surveillance .Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.

D-12

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE 4. 11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT MINIMUM OF LIQUID RELEASE TYPE SAMPLING ANALYSIS TYPES OF ACTIVITY DETECTION FREQUENCY FREQUENCY ANALYSIS (LZ,D) 0>

()>Ci/ml )

Batch Waste Release Tanks<~>

P P Gamma 5x10 1 Waste Monitor Each Batch Each Batch Emi t te rs <'>

Tanks I-131 1x10 ~

b. Waste Evaporator P Dissolved and 1x10 ~

Condensate One Batch/M Entrained Gases Tanks (Gamma Emitters)

Secondary P Composite"'rincipal M H-3 lx10 Waste Sample Each Batch Composite"'

Tank ">

Gross Alpha 'x10-'x10 Treated Laundry and P Sr-89, Sr-90 Hot Shower Each Batch Tanks Fe-55

'x10

~

Continuous Releases+>"'ooling Tower W Principal Gamma 5x10 >

Weir osite'""'

Tank"'ontinuous'" Com Emitters"'issolved Secondary M(7) and Waste Sample Grab Sample Entrained Gases lxlo-'x10 (Gamma Emitters)

I-131 Continuous'" H-3 s

'x10 Composite"

Gross Al ha lx10 >

Continuous'" Sr-89, Sr-90 5x10 Composite'" <"

Fe-55 ~

'x10 D-13

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE 4.11-1 (Continued)

TABLE NOTATZONS (1) The LLD is defined, for purposes of these Operational Requirements, as the smallest concentration of radioactive material i.n a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which may'nclude radiochemical separation:

4.66 sb LLD E V 2.22 x 10~ Y exp (-XDt)

Where:

LLD = the "a priori" lower limit of detection (microCurie per unit mass or volume),

Sb the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute),

E = the counting efficiency (counts per disintegration),

V the sample size (units a'f mass or volume),

2.22 x 10' the number of disintegrations per minute per microCurie, the fractional radiochemical yield, when applicable, the radioactive decay constant for the particular radionuclide (sec '), and lt,t = the elapsed time between the midpoint of sample collection and the time of counting (sec) .

Typical values of E, V, Y, and b,t should be used in the calculation.

rt should be recognized that the LLD is defined as an riori (betore

~a the fact) limit representing the capability of a measurement system measurement.

(2) A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed by a method described in the ODCM to assure representative sampling.

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE 4.11-1 (Continued)

TABLE NOTATIONS (Continued)

(3) The principal gamma emitters for which the LLD operational Requirement applies include the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, and Ce-141. Ce-144 shall also be measured but with a LLD of 5 x 10 '. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent'Release Report pursuant to ODCM, Appendix F, Section F.2 in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974.

(4) A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released.

(5) A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release.

(6) To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream.

Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the .composite sample to be representative of the effluent release.

(7) These points monitor potentia3, release pathways only and not actual release pathways. The potential contamination points are in the Normal Service Water (NSW) and Secondary Waste (SW) Systems. Action under this Operational Requirement is as follows:

a) If the applicable (NSW or SW) monitors in Table 3.3-12 are OPERABLE and not in alarm, then no analysis under this Operational Requirement is required but weekly composites will be collected.

b) If the applicable monitor is out of service, then the week).y analysis for principal gamma emitters will be performed.

c) If the applicable monitor is in alarm or theif principal gamma emitter analysis indicates the presence of radioactivity as defined in the ODCM, then all other analyses of this Operational Requirement shall be performed at the indicated frequency as long as the initiating conditions exist.

(8) The Secondary Waste System releases can be either batch or continuous.

The type of sample required is determined by the mode of operation being used.

I D-15

Shearon Harris Nuclear Power plant (sHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.2 Dose OPERATIONAL REQUIREMENT 3.11.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to UNRESTRICTED AREAS (see Technical Specification Figure 5.1-3}

shall be limited:

a 0 During any calendar quarter to less than or equal to 1.5 mrems to the whole body and to less than or equal to 5 mrems to any organ, and

b. During any calendar year to less than or equal to 3 mrems to the whole body and to less than or equal to 10 mrems to any organ.

APPLICABILITY: At all times.

ACTION:

With the calculated dose from the release of radioactive materials in liquid effluents ex@ceding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the. cause(s} for exceeding the limit(s) and defines the corrective-.actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that. subsequent releases will be 'in compliance with the above limits.

SURVEILLANCE REQUIREMENTS 4.11.1.2 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in. accordance with the methodology and parameters in the ODCM at least once per 31 days.

Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.

D-16

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3/4.11.1 LIQUID EFFLUENTS 3/4. 11.1.3 Li uid Radwaste Treatment S stem OPERATIONAL REQUIREMENT 3.11.1.3 The Liquid Radwaste Treatment System shall be OPERABLE and appropriate portions of the system shall be used to reduce releases of radioactivity when the projected doses, due to the liquid effluent, to UNRESTRICTED AREAS (see Technical Specification Figure 5.1-3) would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31-day period.

APPLICABILITY: At all times.

ACTION:

With radioactive liquid waste being discharged without treatment and in excess of the above limits and any portion of the Liquid Radwaste Treatment System not in operation, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that includes the following information:

Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems,'nd the reason for the inoperability, Action(s) taken to restore the inoperable equipment to OPERABLE status, and

3. Summary description of action(s) taken to prevent a recurrence. f SURVEILLANCE REQUXREMENTS 4.11.1.3.1 Doses due to liquid releases to UNRESTRXCTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when Liquid Radwaste Treatment Systems are not being fully utilized.

4.11.1.3.2 The installed Liquid Radwaste Treatment System shall be considered OPERABLE by meeting Operational Requirements 3.11.1.1 and 3.11.1.2.

Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.

D-17

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 Dose Rate OPERATIONAL REQUIREMENT 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Technical Specification Figure 5.1-1) shall be limited to the following:

For noble gases: Less than or equal to 500 mrems/yr to the whole body and less than or equal to 3000 mrems/yr to the skin, and For Iodine-131, for Iodine-133,'for tritium, and for all radionuclides in particular form with half-lives greater than 8 days: Less than or equal to 1500 mrems/yr to any organ.

APPLICABILITY: At all times.

ACTION:

With the dose rate(s) exceeding the above limits, immediately restore the release rate to within the above limit(s).

SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM.

4.11.2.1.2 The dose rate due to Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits iq accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2.

Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance i nterval .

D-18

TABLE 4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM MINIMUM LOWER LIMIT OF Q Ch SAMPLING ANALYSIS DETECT10N (LLD)( ) tt) 0 lA 0)

GASEOUS RELEASE TYPE FREQUENCY FREQUENCY TYPE OF ACTIVITY ANALYSIS (pCi/ml) r* A rt 0 Q

1. Was te Gas torage P P Principal Gamma Emitters ~

lxlo '

S Tank Each Tank Each Tank 0 C)

Grab Sample

2. Containment Purge P Principal Gamma Emitters(~) lxlO 4 O 0) or Vene'") Each PURGE(3) Each PURGE h 2e C h Grab Sample H-3 (oxide) lxlO s Q

3. a. Plane Vent M(3),(4),(5) Principal Gamma Emiteers lxlO ~

r0ft al5 Stack s R o

Grab Sample H-3 (oxide) lx10 C (D

b. Turbine Bldg M Principal Gamma Emitters lx10 ~

Vent Stack, Grab Sample t r Waste Proc. 0 tt Bldg. Vent H-3 (oxide) (Turbine Bldg. lx10 aO~

Stacks 5 & 5A Vent Stack)

4. All Release Types Continuous( W(7) I-131 lxlo '~

as listed in 1.,2.,

and 3. above Charcoal I-133 lxlo '0 (8), (s), (10) Sample Coneinuous w(>> Principal Gamma Emitters(~) lxlo->>

Particulate Sample Continuous M Gross Alpha lxlo->>

Composite Particulate Sample Continuous Q Sr-89, Sr-90 lxlO "

Composite Particulate Sample OQ C

Sheazon Harris Nuclear Powez Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE 4.11-2 (Continued)

TABLE NOTATlONS (1) The LLD is defined, for purposes of these Operational Requirements, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "zeal" signal.

For a particular measurement system, which may include radiochemical separation:

4. 66 sb LLD =

E v 2.22 gd 10 . Y exp (-hht)

Where:

LLD = the "a priori" lower limit of detection (microCurie 'per unit mass or volume),

Sb the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute),

E = the counting efficiency (couqts pez disintegration),

V = the sample size (units of mass or volume),

2.22 x 10' the number of .disintegrations per minute per microCurie, Y = the fractional radiochemical yield, when applicable, the radioactive decay constant for the particular zadionuclide (sec ') and ht = the elapsed time between the midpoint of sample collection and the time of counting (sec).

Typical values of E, V, Y, and 6t should be used in the calculation.

Xt should be recognized that the iLD is defined as an ~ariori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (atter the fact) limit for a particular measurement.

D-20

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE 4.11-2 (Continued)

TABLE NOTATIONS (Continued)

(2) The principal gamma emitters for which, the LLD Operational Requirement applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, I-131, Cs-134, Cs-137, Ce-141, and Ce-144 in Iodine and particulate releases.

This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent Release Report pursuant to ODCM, Appendix F, Section F.2 in the format outlined in Regulatory Guide 1.21, Appendix,B, Revision 1, June 1974.

(3) Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15$ of RATED THERMAL POWER within a 1-hour period.

(4) Tritium grab samples shall be taken at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the refueling canal is flooded.

(5) Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool.

(6) The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dbse or dose rate calculation made in accordance with Operational Requirements 3.11.2.1, 3.11.2.2, and 3.11.2.3.

(~i Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing, or after removal from sampler.

Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least 7 days following each shutdown, startup, or THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour period and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement does not apply if: (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the reactor coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.

(8) Continuous sampling of Waste Gas Decay Tank (WGDT) releases can be met using the continuous samplers on Wide Range Gas Monitor RM-*1WV-3546-1 on Waste Processing Building Vent Stack 5.

(9) Continuous sampling of containment atmosphere for (1) Venting, (2) Normal Purge, and (3) Pre-entry purge operations, required by Operational Requirement 4.11.2.1.2, can be met using the continuous samplers on Wide Range Gas Monitor RM-01AV-3509-1SA on Plant Vent Stack l.

D-21

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE 4.11-2 (Continued)

TABLE NOTAT1ONS (Continued)

(10) The requirement to sample the containment atmosphere prior to release for normal and pre-entry containment purge operations (that is, to "permit" the release per the ODCM) is required on initial system staztup, and prior to system restart following any system shutdown due to radiological changes in the containment (e.g. valid high alarms on leak detection or containment area monitors) . System shutdown occurring on changes in containment pressure, equipment malfunctions, operational convenience, sampling, and so forth, do not require new samples or release permits.

D-22

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.2 Dose - Noble Gases OPERATIONAL REQUIREMENT 3:11.2.2 The air dose due to noble gases released in gaseous effluents to areas at and beyond the SITE BOUNDARY (see Technical Specification Figure 5.1-3) shall be limited to the following:

During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation, and During any calendar year: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.

APPLICABILITY: At all times.

ACTION:

With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions th'at have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

SURVEILLANCE REQUIREMENTS 4.11.2.2 Cumulative dose contriQutions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.

D-23

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 GASEOUS EFFLUENTS 3/n.11.2.3 Dose Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form OPERATIONAL REQUIREMENT 3.11.2.3 The dose to a MEMBER OF THE PUBLIC from Zodine-131, Zodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released to areas at and beyond the SITE BOUNDARY (see Technical Specification Figure 5.1-3) shall be limited to the following:

During any calendar quarter: Less than or equal to 7.5 mrems to any organ, and

b. During any calendar year: Less than or equal to 15 mrems to any organ.

APPLICABILITY: At all times.

ACTION:

With the calculated dose, from the release of Iodine-131, Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Speci.al Report that identifies the cause(s) for exceeding the limit(s) arid defines the corrective actions that have been taken to reduce the releases and the proposed corrective aMions to be taken to assure that subsequent releases will be in compliance with the above limits.

SURVE1LLANCE REQUIREMENTS 4.11.2.3 Cumulative dose contributions for the current calendar quarter and current calendar year for Zodine-131, Zodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.

D-24

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3/4.11.2 GASEOUS EFFLUENTS 3/4. 11.2.4 Gaseous Radwaste Treatment S stem OPERATIONAL REQUIREMENT 3.11.2.4 The VENTILATION EXHAUST TREATMENT SYSTEM and the GASEOUS RADWASTE TREATMENT SYSTEM shall be OPERABLE and appropriate portions of these systems shall be used to reduce releases of radioactivity when the projected doses in 31 days due to gaseous effluent releases to areas at and beyond the SITE BOUNDARY (see Technical Specification Figure 5.1-3) would exceed:

0.2 mrad to air from gamma radiation, or

b. 0. 4 mrad to. air from beta radiation, or C. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.

APPLICABILITY: At all times.

ACTION'ith radioactive gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that includes the following information:

1. Identification of any inoperable equipment or subsystems, and the reason for the inoperability, Action(s) taken to restore the inoperable equipment to OPERABLE status, and Summary description of action(s) taken to prevent a recurrence.

SURVEILLANCE REQUIREMENTS 4.11.2.4.1 Doses due to gaseous releases to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days in accordance with the methodology and Parameters in the ODCM when the GASEOUS RADWASTE TREATMENT SYSTEM is not being fully utilized.

4.11.2.4.2 The installed VENTILATION EXHAUST TREATMENT SYSTEM and GASEOUS RADWASTE TREATMENT SYSTEM shall be considered OPERABLE by meeting Operational Requirements 3.11.2.1 and 3.11.2.2 or 3.11.2.3.

Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.

D-25

Shearon Harris Nuclear Power plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3/4.11.4 TOTAL DOSE OPERATIONAL REQUIREMENT 3.11.4 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.

APPLICABILITY: At all times.

ACTION:

a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Operational Requirement 3.11.1.2a., 3.11.1.2b., 3.11.2.2a., 3.11.2.2b., 3.11.2.3a.,

or 3.11.2.3b., calculations shall be made including direct radiation contributions from the units and from outside storage tanks to determine whether the above limits of Operational Requirement 3.11.4 have been exceeded. Zf such is the case, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR 20.405(c}, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered. by th5.s report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. Zf the estimated dose(s} exceeds the above limits, and if the release condition resulting in viol'ation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accozdance with the provisions of 40 CFR Part 190.

Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

SURVEILLANCE REQUIREMENTS 4.11.4.1 Cumulative dose contributions fzom liquid and gaseous effluents shall be determined in accordance with Operational Requirements 4. 11. 1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the methodology and parameters in the ODCM.

4.11.4.2 Cumulative dose contributions from direct radiation from the units and from radwaste storage tanks shall be determined in accordance with the methodology and parameters .in the ODCM. This requirement is applicable only under conditions set forth in ACTION a. of Operational Requirement 3.11.4.

Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.

D-26

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 D. 3 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM OPERATIONAL REQUIREMENT 3.12.1 The Radiological Environment Monitoring Program shall be conducted as specified in Table 3.12-1.

APPLICABILITY: At all times.

ACTION:

With the Radiological Environmental Monitoring Program not being conducted as specified in Table 3.12-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by ODCM, Appendix F, Section F.1, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose* to a MEMBER OF THE PUBLIC is less than the calendar year limits of Operational Requirements 3.11.1.2. 3.11.2.2, or 3.11.2.3.

When more than one of the zadionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:

concentration (1) concentration (2) reporting level (1) reporting level (2)

When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted the potential annual dose* to a MEMBER OF THE PUBLIC from all if radionuclides is equal to or greater than the calendar year limits of Operational Requirement 3.11.1.2, 3.11.2.2, or 3.11.2.3. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report required by ODCM, Appendix F, Ssection F.l.

• The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.

D-27

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3/4.12.1 MONITORING PROGRAM OPERATIONAL REQUIREMENT ACTION (Continued)

c. With milk or fresh leafy vegetation samples unavailable from one or more of the sample locations required by Table 3.12-1, identify specific locations for obtaining replacement samples and add them within 30 days to the Radiological Environmental Moni'toring Program given in the ODCM. The specific locations from which samples were unavailable may then be deleted from the monitoring program. Pursuant to ODCM, Appendix F., Section F.2, submit in the next Annual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure(s) and table for the ODCM reflecting the new with supporting information identifying the cause of the 'ocation(s) unavailability of samples and justifying the selection of the new location(s) for obtaining samples.

SURVEILLANCE REQUIREMENTS

4. 12. 1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the specific locations given in the table and figure(s) in the ODCM, and shall be analyzed pursuant to the requirements of Table 3.12-1 and the idetection capabilities required by Table 4.12-1.

Each Surveillance Requi'rement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25~a of the specified surveillance interval.

D-28

TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM* 0 Ol

& 0Ql N

EXPOSURE PATHWAY NUMBER OF REPRESENTATIVE SAMPLING AND TYPE AND FREQUENCY V' AND/OR SAMPLE SAMPLES AND SAMPLE LOCATIONS COLLECTION FREQUENCY OF ANALYSIS rT 0 8

1. Direct Radiation"'orty routine monitoring stations Quarterly. Gamma dose quarterly. ax 0

either with two or more dosimeters or with one instzument for (D measuring and recording dose rate O Ql continuously, placed as follows: k 0 c 4

An inner ring of stations, one in 0 HV each meteorological sector .in the N ID QP general area of the SITE BOUNDARY; V 0

outer ring of stations, one in 0 An Z lD8 each meteorological sector in the DP 6- to 8-km range from the site; The balance of the stations and rr C

to be placed in special interest 0 rt U

areas such as population centers, g, Vl nearby residences, schools, and in one or two areas to serve as control stations.

TABLE 3. 12-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY NUMBER OF REPRESENTATIVE SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE SAMPLES AND SAMPLE LOCATIONSni COLLECTION FREQUENCY OF ANALYSIS

2. Airborne Samples from five locations: Continuous sampler Radioiodine Cannister:

Radioiodine and operation with sample I-131 analysis weekly.

Particulates collection weekly, or Three samples the three SITE from close to BOUNDARY more frequently required by dust if locations, in different loading. Particulate Sam ler:

sectors, of the highest Gross beta radioactivity calculated annual average analysis following ground-level D/Q; filter change; "'nd gamma isotopic vicinity analysis"'f One sample from the composite (by location) of a community having the quarterly.

highest calculated annual average ground-level D/Q; and One sample from a control location, as for example 15 to 30 km distant and in the aI least prevalent wind direction.

o

3. Waterborne

a. Surface One sample upstream. Composite sample over Gamma isotopic 1-month period." Composite for analysis"'onthly.

One sample downstream. tritium analysis quarterly.

b. Ground Samples from one or two Quarterly. Gamma isotopic'" and sources only to be if affected'n.

likely tritium analysis quarterly.

TABLE 3.12-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM O III th &

EXPOSURE PATHWAY NUMBER OF REPRESENTATIVE SAMPLING AND TYPE AND FREQUENCY th O AND/OR SAMPLE SAMPLES AND SAMPLE LOCATIONSIII COLLECTION FREQUENCY, OF ANALYSIS rt 0

3. Waterbourne S (Continued) tI A 0

c. Drinking One sample in the vicinity Composite sample over I-131 analysis on each Ol 0

of the nearest downstream 2-week period<'I when composite when the dose ne municipal water supply intake I-131 analysis is calculated for the VZ form the Cape Fear River. performed; monthly consumption of the n c composite otherwise. water is greater than c 0 I I One sample from a control 1 mrem per (D location. for gross year."'omposite CT Q beta and gamma isotopic 0 analysesI" monthly. 0 Composite for tritium Z 88 analysis quarterly.

d. Sediment one sample in the vicinity of the Semiannually. Gamma isotopic VI Ql from cooling tower blowdown discharge analysisI'I semiannually. On Shoreline- in an area with existing or potential recreational value. n-U Q M

4. 1ngestion 'z

a. Milk Samples from milking animals in semimonthly when Gamma isotopic"'nd three locations within 5 km animals are on I-131 analysis distance having the highest dose pasture; monthly semimonthly when potential. If there are none, at other times. animals are on pasture; then one sample from milking monthly at other times.

animals in each of three areas between 5 to 8 km distant where doses are calculated to be greater than 1 mrem per yr.<'I One sample from milking animals at a control location 15 to 30- km distant and in the least prevalent wind direction.

TABLE 3.12-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PAT%'/AY NUMBER OF REPRESENTATIVE SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE SAMPLES AND SAMPLE LOCATIONS~u COLLECTION FREQUENCY OF ANALYSIS

4. Ingestion (Continued)

b. Fish and One sample of Sunfish, Catfish, and Sample in season or Gamma is o topic Invertebrates Large-Mouth Bass species in vicinity semiannually if they analysis<" on of plant discharge area. are not seasonal. edible portions.

One sample of same species in areas not influenced by plant discharge.

c. Food Products samples of three different kinds Monthly during Gamma isotopic"'nd of broad leaf vegetation grown growing season. I-131 analysis.

nearest each of two different.

offsite locations of highest predicted annual average ground level D/Q if not performed.

milk sampling is 1

One sample of each of the similar Monthly during Gamma isotopic<" and broad leaf vegetation grown 15 to growing season. I-131 analysis.

30 km distant in the least prevalent wind direction if milk sampling is not performed.

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE 3.12-1 (Continued)

TABLE NOTATIONS (1) Specific parameters of distance and direction sector from the centerline of one reactor, and additional description where pertinent, shall be provided for each and every sample location in Table 3.12-1 in a table and figure(s) in the ODCM.

Refer to NURE6-0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," October 1978, and to Radiological Assessment Branch Technical Position, Revision 1, Novembez 1979. Deviations aze permitted from the required sampling schedule if specimens aze unobtainable due to circumstances such as hazardous conditions, seasonal unavailability, and malfunction of automatic sampling equipment. If specimens are unobtainable due to sampling equipment malfunction, effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.3. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the Radiological Environmental Monitoring Program. Pursuant to ODCM, Appendix F, Section F.2, submit in the next Annual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure(s) and table for the ODCM reflecting the new location(s) with supporting information identifying the cause of the unavailability of samples for that pathway and justifying the selection of the new location(s) for obtaining samples.

(2> One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring dizect radiation. (The 40 stations are not an absolute number. The number of direct radiation monitoring stations may be reduced according to geographical limitations; e.g., at an ocean site, some sectors will be over water so that the number of dosimeters may be reduced accordingly. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information within minimal fading.)

(3) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

D-33

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE 3.12-1 (Continued)

TABLE NOTATIONS (Continued)

Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility. I (5) The "upstream sample" shall be taken at a distance beyond significant influence of the discharge. The downstream" sample shall be taken in an area beyond but near the mixing zone. "Upstream" samples in an estuary must be taken far enough upstream to be beyond the plant influence. Salt water shall be sampled only when the receiving water is utilized for recreational activities.

(6) A composite sample is one in which the quantity (aliquot) of liquid sampled is proportional to the quantity of flowing liquid and in which the method of sampling employed results in a specimen that is representative of the liquid flow. In this program composite sample aliquots shall be collected at time intervals that are very short (e.g., hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample.

Groundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination.

The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM.

D-34

TABLE 3.12-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES 0 Ol tb 0 Ol Ql WATER AIRBORNE PARTICULATE FISH MILK FOOD PRODUCTS rt 0 ANALYSIS (pci/1) OR GASES (pCi/m') (pCi/k, wet) (pCi/1) (pCi/k , wet) 6 H-3 20, 000* e0 x Ql 0

Mn-54 1, 000 30, QQQ n rn Fe-59 400 10, 000 in 0 c I

Co-58 1, 000 30, 000 0 Ql I

0 Co-60 300 10, 000 0

Z Z Zn-65 300 20, 000 Q O C

Zi-Nb-95 400 I I I-131 0.9 100 0 rt U

n~

ZM Cs-134 30 10 1 000 60 1, 000 z'Q Cs-137 50 20

• 2,000 70 2, 000 Ba-La-140 200 300

• For drinking water samples: This is 40 CFR Part 141 value. -1f no drinking water pathway exists, a value of 30,000 pCi/1 may be used.

• ~If no drinking water pathway exists, a value of 20 pci/1 may be used.

DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS O Ol th K W 0 LIMIT OF DETECTION LLD 'OWER PP rt 0 0

WATER AIRBORNE PARTICULATE FISH MILK FOOD PRODUCTS SEDIMENT U X ANALYSIS (pCi/1) OR GASES (pCi/m') (pCi/kg, wet) (pCi/1) (pCi/kg, wet) (pci/kg, dry) 0 N tj Gross Beta 0.01 V' th H-3 2000* rz n a rn I

Mn-54 130 (D rrt0 Pn Fe-59 30 260 0

Z oZ Co-58, 60 130 Zn-65 30 260 l Zr-Nb-95 O ct D

O~V) 1-131 0.07 60 g r

Cs-134 0. 05 130 ~ 15 60 150 Cs-137 18 0. 06 150 18 80 180 Ba-La-140 15

• If no drinking water pathway exists, a value of 3000 pci/1 may be used.

<<<<If no drinking water pathway exists, a value of 15 pCi/1 may be used.

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE 4.12-1 (Continued)

TABLE NOTATIONS (1) This list does not mean that only'these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to ODCM, Appendix F, Section F.l.

(2) Required detection capabilities for thezmoluminescent dosimeters used for environmental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13.

(3) The LLD is defined, for purposes of these Operational Requirements, as the smallest concentzation of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation:

4.66 sb LLD E V 2.22 Y exp (-Mt)

Where:

LLD the "a priori" lower limit ofI detection (picoCurie per unit mass or volume),

Sg the standard deviation of the backgiound counting rate or of the counting rate of a blank sample as appropriate (counts per minute),

the counting efficiency (counts per disintegration),

V the sample size (units of mass or volume),

2.22 = the number of disintegrations per minute pez picoCurie, the fractional radiochemical yield, when applicable, the radioactive decay constant for the particular zadionuclide (sec'), and the elapsed time between environmental collection, or end of the sample collection period and the time of counting (sec).

Typical values of E, V, Y, and 6t should be used in the calculation.

D-37

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE 4.12-1 (Continued)

TABLE NOTATIONS (Continued)

It should be recognized that the LLn is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to ODCM, Appendix F, Section F.l.

D-38

Shea ron Harris Nuclear Pcwer Plant (SHNPP) August 1995 Of fsite Dose Calculation Manual (ODCM) Rev. 6 3/4.12.2 LAND USE CENSUS OPERATIONAL REQUIREMENT 3.12.2 A Land Use Census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence, and the nearest garden* of greater than 50 m~ (500 ft~)

producing broad leaf vegetation.

APPLICABILITY: At all times.

ACTXON With a Land Use Census identifying a location(s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Operational Requirement 4.11.2.3, pursuant to ODCM, Appendix F,. Section F.2, identify the new location(s) in the next Annual Radioactive Effluent Release 'Report.

With a Land Use Census identifying a location(s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20%

greater than at a location from which samples are currently being obtained in accordance with Operational Requirement 3.12.1, add the new location(s) within 30 days to the Radiological Environmental Monitoring Program given in the ODCM. The sampling location(s), excluding the control station location, having the lowest calculated dose or dose commitment(s), via the same exposure pathway, may be deleted from this monitoring program after October 31,df the year in which this Land Use Census was conducted. Pursuant to ODCM, Appendix F, Section F.2, submit in the next Annual Radioactive Effluent Release Report documentation for a chahge in the ODCM including a revised figure(s) and table(s) for the ODCM ieflecting the new location(s) with information supporting the change in sampling locations.

SURVEILLANCE REQUIREMENTS 4.12.2 The Land Use Census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities. The results of the Land Use Census shall be included in the Annual Radiological Environmental Operating Report .pursuant to ODCM, Appendix F, Section F.l.

Each Surveillance Requirement shall be performed within the specified survei'llance intezval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.

• Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census. Operational Requirements broad leaf vegetation sampling in Table 3.12-1, Part 4.c., shall be followed, for including analysis of control samples.

D-39

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM OPERATIONAL REQUIREMENT 3.12.3 Analyses shall be performed on all radioactive materials, supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission, that correspond to samples required by Table 3.12-1.

APPLICABILITY: At all times.

ACTION:

With analyses not being performed as required 'above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to ODCM, Appendix F, Section F.l, SURVEILLANCE REQUIREMENTS 4.12.3 The Interlaboratory Comparison Program shall be described in the ODCM.

A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to ODCM, Appendix F, Section F.l.

Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.

D-40

Shearon Harris Nuclear Power plant (SHNPP) August 19g5 Offsite Dose Calculation Manual (ODCM) Rev. 6 APPENDIX E PROGRAMMATIC CONTROL BASES The Bases for the ODCM Operational Requirements are detailed in Sections:

E.l - Instrumentation E.2 Radioactive Effluents E.3 - Radiological. Environmental Monitoring

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 E.l INSTRUMENTATION BASES 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.10 Radioactive Li uid Effluent Monitorin Instxumentation The radioactive liquid effluent instrumentation is provided to monitox and control, as applicable, the releases of radioactive materials in liquid effluents duxing actual or potential releases of liquid effluents. The Alarm/Trip Set Points for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm/trip will occur pxior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Critex'ia 60, 63, and 64 of Appendix A to 10 CFR Paxt 50.

3/4.3.3.11 Radioactive Gaseous Effluent Monitorin Instrumentation The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm/Trip Set Points for these instruments shall be calculated and adjusted in accordance with the methodology and paxameters in the ODCM to ensure that the alaxm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 -of Appendix A to 10 CFR Part 50. The sensitivity of any noble gas activity monitoxs used to show compliance wj.th the gaseous effluent release requirements of Operational Requirement 3.11.2.2 shallbe such that concentrations as low as 1 x 10

'Ci/ml are measurable.

E-2

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 E.2 RADIOACTIVE EFFLUENTS BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 Concentration This Operational Requirement is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table 2, Column 2.

This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within: (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR Part 20.106(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption .that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (XCRP) Publication 2.

The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLDg and other detection limitsg can be found in HASL Procedures Manual, HASL-300 (revised annually),

Currie, L. A., "Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry, " Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., "Detection Limits for Radioanalytical Counting Techniques, " Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

3/4.11.1.2 Dose This Operational Requirement is provided to implement the requirements of Sections XI.A, IIX.A., and IV.A of Appendix X, 10 CFR Part 50. The Operational Requirement implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section XV.A of Appendix I to assure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable". The dose calculation methodology and parameters in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.

E-3

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 E.2 RADXOACTIVE EFFLUENTS BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.2 Dose (continued)

The equations specified i.n the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, " Revision 1, October 1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Xmplementing Appendix I," April 1977.

3/4.11.1.3 Li uid Radwaste Treatment S stem The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This Operational Requirement implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in. Section IZED of Appendix I to 10 CFR Part 50.

The specified limits governing the use of appropriate portions of the Liquid Radwaste Treatment System were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50 for liquid effluents.

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 E.2 RADIOACTIVE EFFLUENTS BASES 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 Dose Rate This Operational Requirement is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column Z. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 [10 CFR .Part 20.106(b)] . For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY.

Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the ODCM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrems/year to the whole body or to less than or equal to 3000 mzems/year to the skin.

These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/year.

The required detection capabilities for radioactive material in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually)g Currie, L. A., "Limits for Qualitative Detection and Quantitative Determination Appliaation to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

E-5

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 E.2 RADZOACTXVE EFFLUENTS BASES 3l4.11.2 GASEOUS EFFLUENTS 3/4.11.2.2 Dose - Noble Gases This Operational Requirement is provided to implement the requirements of Section XZ.B, IXZ.A, and IV.A of Appendix I, 10 CFR Part 50. The Operational Requirement implements the guides set forth in Section I.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section XV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section ZZZ.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures. based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Dose to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, " Revision I, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors, "

Revision 1, July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 E.2 E RADIOACTIVE EFFLUENTS BASES 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.3 Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form This Operational Requirement is provided to implement the requirements of Sections II.C, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Operational Requirements are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section ZV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonable achievable." The ODCM calculational methods specified in the Surveillance Requirements implement the requirements in Section IXZ.A of Appendix X that conformance with the guides of Appendix X be shown by calculational proceduzes based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.

The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispezsion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors, "

Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate Operational Requirements foz Xodine-131, Iodine-133, tritium, and radionuclides in particulate form with half-lives greater, than 8 days are dependent upon the existing radionuclide pathways to man in the areas at and beyond the SITE BOUNDARY,. The pathways that were examined in the development of the calculations were: (1) individual inhalation of airborne radionuclides, (2) deposition of zadionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and (4) deposition of the ground with subsequent exposure of man.

E-7

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6

"-.2 E RADIOACTIVE EFFLUENTS, BASES (continued) 3/4.11.2 GASEOUS EFFLUENTS 3/4. 11.2.4 Gaseous Radwaste Treatment S stem The OPERABILITY of the WASTE GAS HOLDUP SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensure that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This Operational Requirement implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design given in Section ~ .D of Appendix I to 10 CFR Part 50. The

'bjectives specified limits governing the use of, appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 E.2 RADIOACTIVE EFFLUENTS BASES (continued) 3/4.11.3 SOLID RADIOACTIVE WASTES This specification implements the requirements of 10 CFR 50 36a 10 CFR 61'nd General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to, waste type, waste pH, waste/liquid/SOIIDIFICATION agent/catalyst ratios, waste oil content, waste principal chemical constituents, and mixing and curing times.

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 E.2 RADIOACTIVE EFFLUENTS BASES (continued) 3/4.11.4 TOTAL DOSE This Operational Requirement is provided to meet the dose limitations of 10 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The Operational Requirement requires the preparation and submittal of a Special Report whenever the calculated doses due to releases of radioactivity and to radiation from uranium fuel cycle sources exceed 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to four reactors, it is highly unlikely that, the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the units and from outside storage tanks are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that, the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Operational Requirements 3.11.1.1 and 3.11.2.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

E-10

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM} Rev. 6 ED 3 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.1 MONITORING PROGRAM The Radiological Environmental Monitoring Program required by this Operational Requizement provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest Potential radiation exPosure of MEMBERS OF THE PUBLIC resulting from the plant operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the Radiological Effluent Monitoring Program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience.

The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LlDs) . The LLDs required by Table 4.12-1 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an ~ariori (before the fact) limit representing the capability of a particular measurement.

Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., "Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry, " Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., "Detection Limits for Radioanalytical Counting Techniques". Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

3/4.12.2 LAND USE CENSUS This Operational Requirement is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the Radiological Environmental Monitoring Program are made, if required, by the results of this census. - The best information from the door-to-door survey, from aerial survey, or from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 50 m~ provides assurance that significant exposure pathways via leafy vegetables will be i.dentified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To dete mine this minimum garden size, the following assumptions were made: (1) 20pe of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m,.

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. a

"-.3 RADIOLOGICAL ENVXRONMENTAL MONITORING BASES (continued) 3/4.12.3 XNTERLABORATORY COMPARXSON PROGRAM The requirement f'r participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part" of the quality assurance program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section XV.B.2 of Appendix r I to 10 CFR Part 50.

E-12

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 APPENDIX F ADMINISTRATIVE CONTROLS The Reporting Requirements pertaining to the ODCM Operational Requirements are detailed in Sections:

F.l - Annual Radiological Environmental Operating Report F.2 - Annual Radioactive Effluent Release Report F.3 Major changes to the Radwaste Treatment System (liquid and gaseous)

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6

.i Annual Radiolo ical Environmental 0 eratin (Formerly part of Specification 6.9.1.3)

Re ort Routine Annual Radiological Environmental Operating Reports, covering the operation of the unit during the previous calendar year, shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following initial criticality.

The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, with operational controls, as appropriate, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment.

The reports shall also include the results of the Land Use Census required by Operational Requirement 3.12.2.

The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the OFFSITE DOSE CALCULATION MANUAL, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining, the reasons for the missing results.

The missing data shall be submitted as soon as possible in a supplementary report. The reports shall also include the ~following: a summary description of the Radiological Environmental Monitoring Program; at least two legible all sampling locations keyed to a table giving distances and maps'overing.

directions from the centerline'of the reactor; the results of licensee participation in the Interlaborato'ry Comparison Program and the corrective action taken if the specified program is not being performed as required by Operational Requirement 3.12.3; reasons for not conducting the Radiological Environmental Monitoring Program as required by Operational Requirement 3.12.1, and discussion of all deviations from the sampling schedule of Table 3.12-1; discussion of environmental sample measurements that exceed the reporting levels of Table 3.12-2 but are not the result of plant effluents, pursuant to ACTION b. of Operational Requirement 3.12.1; and discussion of all analyses in which the LLD required by Table 4.12-1 was not achievable.

One map shall cover stations near the EXCLUSION AREA BOUNDARY; a second shall include the more distant station.

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 Annual Radioactive Effluent Release Re ort (Formerly part of Specification 6.9.1.4)

Routine Annual Radioactive Effluent Release Report covering the operation of the unit during the previous 12 months of operation shall be submitted of each year. The period of the first report shall begin with the date byofMay 1 initial criticality.

The Annual Radioactive Effluent Release Report shall include a summary of the quantities of radioactive liquid and gaseous effluents released from the unit as outlined in Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants, " Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

The Annual Radioacpive Effluent Release Report shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability. This report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year.

This report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Technical Specification Figure 5.1-3) during the report period.

All assumptions used in making these assessments; i.e., specific activity, exposure time, and location, shall be included in these reports. The meteorological conditions concurrent wit'h the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

The Annual Radioactive Effluent Release Report shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, "Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Revision 1, October 1977.

In lieu of submission with the Annual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

F-3

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 F.2 Annual Radioactive Effluent Release Re ort (continued)

The Annual Radioactive Effluent Release Report shall include a list and description of unplanned releases, from the site to UNRESTRICTED AREAS, of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Annual Radioactive Effluent Release Report shall include any changes made during the reporting period to the ODCM, pursuant to Technical Specification

6. 14, as well as any major change to Liquid and Gaseous Radwaste Treatment Systems pursuant to ODCM, Appendix F, Section F.3. It shall also include a listing of new locations for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to Operational Requirement 3.12.2.

The Annual Radioactive Effluent Release Report shall also include the following: an explanation as to w'..~ the inoperability of liquid or effluent monitoring instrumentation was not corrected within the timegaseous in Operational Requirement 3.3.3.10 or 3.3.3.11, respectively; and a specified description of the events leading to liquid holdup tanks or gas storage tanks exceeding the limits of Technical Specification 3.11.1.4 or 3.11.2.6, respectively.

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6

" Ma'or Chan es to Li uid and Gaseous Radwaste Treatment part of Specification 6.15)

Licensee-initiated major changes S

stems'Formerly to the Radwaste Treatment Systems (liquid and gaseous):

Shall be reported to the Commission in the Annual Radioactive Effluent Release Report for the period in which the evaluation was reviewed in accordance with Specification 6.5. The discussion of each change shall contain:

A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CER 50.59.

Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information.

3. A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems.

An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents that differ from those previously predicted in the License application and amendments thereto.

An evaluation of the change, which shows the expected maximum exposures, to a MEMBER OF THE, PUBLIC in the UNRESTRICTED AREA and to the general population, that di'ffer from those previously estimated in the License application and amendments thereto.

A comparison of the predicted releases of radioactive materials in liquid'and gaseous effluents to the actual releases for the period prior to when the change is to be made.

An estimate of the exposure to plant operating personnel as a result of the change.

Documentation of the fact that the change was reviewed and found acceptable in accordance with Technical Specification 6.5.

b. Shall become effective upon review and acceptance in accordance with Technical Specification 6.5 Licensees may choose to submit the information called for in the Operational Requirement as part of the annual FSAR update.

Shearon Harris Nuclear Power plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 APPENDIX G DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout the ODCN Operational Requirements.

ACTION ACTION shall be that part of an ODCM Operational Requirement which prescribes remedial measures required under designated conditions.

ANALOG CHANNEL OPERATIONAL TEST An ANALOG CHANNEL OPERATIONM TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, interlock and/or Trip Setpoints such that the Setpoints are within the required range and accuracy.

CHANNEL CALIBRATION A CHANNEL CALXBRATZON shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of, input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

D1GITAL CHANNEL OPERATIONAL TEST A DIGITAL CHANNEL OPERATXONM TEST shall consist of exercising the digital computer hardware using data base manipulation to verify OPERABILXTY of alarm and/or trip functions.

OSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microCurie/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors" used for this calculation shall be those listed in Table XZX of TXD-14844, "Calculation of Distance Factors for Power and Test Reactor Sites."

EXCLUSION AREA BOUNDARY The EXCLUSION AREA BOUNDARY shall be that line beyond which the land is not controlled by the licensee to limit access.

FREQUENCY NOTATION The FREQUENCY NOTATXON specified for the performance of Operational Requirements shall correspond to the intervals defined in Table G-1.

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 DEFINITIONS (continued)

GASEOUS RADWASTE TREATMENT SYSTEM A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system off-gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

MEMB R(S) OF THE PUBLIC MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

CFFSITE DOSE CALCULATION MANUAL The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program.

OPERABLE - OPERABILITY A system, subsystem, txain, component or device shall be or have OPERABILITY when all necessary it attendant instrumentation, controls, electrical OPERABLE is capable of performing its specified function(s),and when power, cooling water, lubrication or other auxiliary'equipment that are required for the ox'eal system,,subsystem, train, component, or device to perform its function(s) are also capable of performing their related support function(s).

CPERATIONAL MODE MODE An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table G-2.

.PROCESS CONTROL PROGRAM The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71 and Federal and State regulations, burial ground requirements, and other requirements governing the disposal of radioactive waste.

PURGE - PURGING PURGE or PURGING shall be any controlled pxocess of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a mannex that replacement air or gas is required to purify the confinement.

G-2

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 DEFINITIONS (continued)

SITE BOUNDARY For these Operational Requirements, the SITE BOUNDARY shall be identical to the EXCLUSION AREA BOUNDARY defined above.

SOLIDIFICATION SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.

SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

UNRESTRICTED AREA An UNRESTRICTED AREA shall be 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, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Features Atmo'spheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING VENTING shall be the controlled process of discharging air or gas from a confinement'to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

G-3

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE G-1 FREQUENCY NOTATION NOTATION FREQUENCY' At least once per 12 hours.

At least once per 24 hours.,

At least once per 7 days.

At least once per 31 days.

At least once per 92 days.

SA At least once per 184 days.

At least once per 18 months.

S/U Prior to each reactor startup.

N.A. Not applicable.

Completed prior to each release.

- Each Surveillance Requirement shall be performed within the specified surveillance

'nterval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.

Shearon Harris Nuclear Power Plant (SHNPP) August 1995 Offsite Dose Calculation Manual (ODCM) Rev. 6 TABLE G-2 OPERATXONAL MODES Mode Reactivity Keff RATED AVERAGE COOLANT Condition THERMAL POWER* TEMPERATURE Power 0 erations x 0.99 > 5% a 350 F Startup w 0.99 s 5% 350'F Hot Standby 0.99 350'.F Hot Shutdown 0.99 350'F > Tavg > 200'F Cold Shutdown 0.99 c 200 F Refueling ** 0.95 s 140 F Excluding decay heat.

Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

G-5

~

g

Attachment 1 FIRE PROTECTION BASELINE PILOT INSPECTION LICENSEE PRE-BRIEF ATTENDEES NAME ORGANIZATION J. Hannon Office of Nuclear Reactor Regulation (NRR)/Division of Systems Safety and Analysis (DSSA)/ Plant Systems Branch (SPLB)

P. Madden NRR/DSSA/SPLB L. Whitney NRR/DSSA/SPLB S. Wong NRR/DSSA/Probabilistic Safety Assessment Branch S. West NRR/DSSA/SPLB F. Emerson Nuclear Energy Institute P. Boulden Virginia Power G. Wiseman Rll M. Branch NRC/NRR R. Oates Preferred Licensing Services, Inc., (PLC)

M. Callahan Government Services Incorporated (GSI)

R. Fuhrmeister Rl P. Milano NRR/DLPM G. Saiamon PSEG F. Yeich PSEG B. Thomas PSEG M. Cumbest Entergy D. Vann State of New Jersey C. Alexander PSEG D. Raleigh Bechtel Power R. Rispoli Entergy T. Raimondo NISYS Corporation T. Mcllaine PSEG C. Cahill Rl S. Trubatch Winston and Strawn F. dePeralta Tri-En Corporation W. Ruland Rl B. McDevitt DE&S K. Erdman OPPD Ft. Calhoun S. Nowlen Sandia National Laboratory K. Mathur PSEG J. Keenon PSEG T. Peterson OPPD D. Frumkin Virginia Power F. Wyant Sandia National Laboratory J. Hyslop NRR/DSSA/SPSB A. Madison US NRC J. McMannis Omaha Public Power

FIRE PROTECTION BASELINE PILOT INSPECTION LICENSEE PRE-BRIEF ATTENDEES (continued)

NAME ORGANIZATION K. Young Rl R.'Nease RIV H. Walker RIV R. Kalentari EPM, Inc.

J. Lechner NPPD - Cooper Nuclear Station S. Burgess Rill C.Johnson RIV D. Modeen Nuclear Energy Institute J. LaChance Sandia National Laboratory

Attachment 2 LICENSEE PRE-BRIEF: FIRE PROTECTION BASELINE PILOT INSPECTIONS Location: Holiday Inn Bethesda 81 20 Wisconsin Avenue Bethesda, Maryland 20814 (301-652-2000)

(South from Medical Center Red Line Metro Stop, North from Bethesda Red Line Metro Stop)

Se tember 15 1999 8:00-8:15 Introduction, Discussion of Agenda and Planned Spring 2000 Post-pilot Workshop John N. Hannon, Chief, Plant Systems Branch, NRR 8:15-8:45 NRC Fire Protection Functional Inspection (FPFI) Program Final Results, Conclusions, and Commission Directed Follow-on Actions (SECY-99-140)

Steven West, Chief, Fire Protection Engineering and Special Projects Section, Plant Systems Branch, NRR 8:45-9:15 Overview of the New Reactor Oversight Program, Oversight Program Significance Determination Process (SDP), and the Baseline Pilot Program Alan Madison, Transition Task Force Leader, NRR 9:15-9:30 Structure and Content of the Fire Protection Baseline Inspection Procedure Leon E. Whitney, Plant Systems Branch, NRR 9:30-9:45 Break 9:45-10:00 Resident Inspector and Triennial Team Preparation, Coordination and Onsite Inspection Activities Leon E. Whitney, Plant Systems Branch, NRR 10:00-11:15 Pilot Plant Presentations on Plant Specific Fire Protection Activities and Initiatives Three Licensee Speakers TBD 11:15-11:45 Open Discussion of the Topics Covered During the Morning Session 11:45-1:00 Lunch Break

1:00-1:30 History of the Development of the Fire Protection Specific SDP ["Fire Protection Risk Significance Screening Methodology" (FPRSSM)]

Patrick M. Madden, SPLB/NRR J.S. Hyslop, SPSB/NRR 1:30-2:30 Practical Examples of Application of the FPRSSM 1

Patrick M. Madden, SPLB/NRR J.S. Hyslop, SPSB/NRR 2:30-3:00 Discussion of Use of FPRSSM Results Patrick M. Madden, SPLB/NRR J.S. Hyslop, SPSB/NRR 3:00-3:15 Break 3:15-4:00 Open Discussion/Question and Answer Session p

4:00 Adjourn

FIRE PROTECTION BASELINE INSPECTION PROCEDURE STRUCTURE AND CONTENT LEON WMITNEY SPLB/DSSA/NRR SEPTEMBER, 1999

. BASES

~ FIRE RISK CAN EXCEED TOTAL RISK FROM INTERNAL EVENTS

~ DEFENSE IN DEPTH

~ PREVENT FIRES

~ RAPIDLY DETECT, CONTROL AND EXTINGUISH FIRES

~ PROTECT P/F SAFE SHUTDOWN CAPABILITY

~ FIRE BARRIERS

~ PROCEDURES

. ~ EQUIPMENT/SYSTEMS

~ ~ NO PERFORMANCE INDICATORS

LEVEL OF EFFORT

~ MONTHLY RESIDENT TOUR OF 2-4 PLANT AREAS FOR 1 HOUR I

~ COMBUSTIBLES/I6 NITION SOURCES

~ FIRE PROTECTION SYSTEMS, EQUIPMENT AND FEATURES

~ MATERIALCONDITION

~ OP E RATIONALSTATUS

~ OPERATIONAL LINEUP

~ FIRE BARRIERS

~ ANNUALFIRE DRILL OBSERVATION

RESIDENT TOUR

~ TRANSIENT COMBUSTIBLES, WELDING, HOT WORK

~ FIRE DETECTOR LOCATIONS, CONDITION, OPS STATUS

~ SPRINKLER SYSTEMS:

LOCATION, SUPPLY, CONDITION 1

~ GASEOUS SYSTEMS: NOZZLE OBSTRUCTIONS, SUPP. AGENT CHARGE, MAN/AUTO MODE, ROOM SEALING, MATERIAL CONDITION

~ EXTINGUISHERS, HOSE STATIONS AND STANDPIPES

RESIDENT TOUR INSPECTION REQUIREMENTS

~ PASSIVE FIRE PROTECTION

~ RACEWAY FIRE BARRIERS

~ FIRE DOOR MATERIAL CONDITION

~ VENTILATIONDAMPEBS

~ . STRUCTURAL STEEL FIRE PROOFING

~ ELECTRICAL FIRE PENETRATION SEALS

~ COMPENSATORY MEASURE ADEQUACY

ANNUALFIRE BRIGADE DRILL OBSERVATION

~ SPECIFIC TOPICS INCLUDE:

~ PROTECTIVE CLOTHING, BREATHING APPARATUS AND FIRE FIGHTING EQUIPMENT

~ FIRE HOSE USE

~ LEADER CONTROL, COMMUNICATIONS; AND FIRE AREA ENTRY

~ SMOKE REMOVAL EFFECTIVENESS

~ USE OF FIRE PRE-PLAN STRATEGIES

TRIENNIALTEAM INSPECTION PREPARATION

~ BRA RISK INSIGHTS REPORT:

~ FIRE AREA RISK RANKINGS

~ CONDITIONALCORE DAMAGE PROBABILITY

~ TRANSIENT SEQUENCES e TEAM LEADER SELECTION OF THREE TO FIVE PLANT AREAS IMPORTANT TO RISK e 2-3 DAY INFORMATION AND ACCESS BADGING SITE VISIT

~ TEAM LEADER INSPECTION PLAN DEVELOPMENT

TRIENNIALTEAM INSP ECTION CONDUCT

~ SAFE SHUTDOWN SYSTEM SELECTION ADEQUACY

~ SYSTEMS SEPARATION EVALUATIONSAGAiNST REQUIREMENTS OF III.G.2 OF APPENDIX R (INCLUDING AREA DETECTION AND SUPPRESSION)

~ FIRE SUPPRESSION DAMAGE ASSESSMENT FOR REDUNDANT TRAINS OF EQUIPMENT

~ FIRE AREA BOUNDARY DESIGN ADEQUACY

TRIENNIALTEAM INSPECTION CONDUCT (CONTINUED)

~ OPERATOR RECOVERY ACTIONS

~ SMOKE REMOVAL

~ DEWATERING

~ CONTROLLED RE-ENERG IZATION

~ RETURN TO SERVICE

~ MANUALFIRE FIGHTING CAPABILITYASSESSMENT (DRILL OBSERVATION)

~ COMP ENSATORY.

MEASURES ADEQUACY

TRIENNIALTEAM INSPECTION CIRCUIT ANALYSIS

~ COMMON POWER SUPPLY CONCERN (INCLUDING MULTIPLE HIGH IMPEDANCE FAULT CONDITIONS)

~ COMMON ENCLOSURE CONCERN (ELECTRICAL FAULT PROTECTION FROM NON-ESSENTIAL CIRCUITS)

~ SPURIOUS SIGNAL CONCERN

~ HOT SHORTS

~, SHORTS TO GROUND

~ OPEN CIRCUITS FUSE/BREAKER COORDINATION

g1 TRIENNIALTEAM INSPECTION ALTERNATIVESHUTDOWN CAPABILITYDESIGN

~ LICENSEE SAFE SHUTDOWN SYSTEM SELECTION ADEQUACY

~ VERIFICATION OF HOT AND COLD SHUTDOWN CAPABILITY WITH AND WITHOUT OFFSITE POWER

~ EFFECT OF FIRE-INDUCED CIRCUIT FAULTS ON TRANSFER OF CONTROL FROM THE CONTROL ROOM TO THE ALTERNATIVESHUTDOWN PANEL

TRIENNIALTEAM INSPECTION ALTERNATIVESHUTDOWN CAPABILITYIMPLEMENTATION

~ OPERATOR TRAINING

~ SHUTDOWN STAFFING (USING ONSITE PERSONNEL ONLY)

~ TIME.REQUIREMENTS, COMMS AND HUMAN FACTORS

~ ALTERNATIVETRANSFER AND CONTROL

~ OPERABILITYTESTING AND INCLUSION IN TECH SPECS

~ COMPENSATORY MEASURES VERIFICATION

TRIENNIALTEAM INSPECTION ALTERNATIVESHUTDOWN CAPABILITYIMPLEMENTATION (CONTINUED)

~ PORTABLE RADIO AND FIXED COMMUNICATIONS

~ AVAILABLE

~ OPERABLE

~ ADEQUATE FOR CONDUCT AND COORDINATION OF ACTIVITIES

~ AMBIENT NOISE LEVELS

~ CLARITYOF RECEPTION

~ RELIABILITY

~ COVERAGE

TRIENNIALTEAM INSPECTION ALTERNATIVESHUTDOWN CAPABILITYIMPLEMENTATION (CONTINUED)

~ EMERGENCY LIGHTING

~ ACCESS AND EGRESS ROUTES, AND CONTROL LOCATIONS

~ PROTECTION OF EMERG.

LIGHTING FROM FIRE LOSS

~ 8 HOUR CAPACITY

~ TESTING AND MAINTENANCE

~ IN-PLANT TESTS OF AIMING AND ILLUMINATIONLEVEL

TRIENNIALTEAM INSPECTION ALTERNATIVESHUTDOWN CAPABILITYIMPLEMENTATION (CONTINUED)

~ COLD SHUTDOWN REPAIRS

~ DAMAGE SPECIFIC REPAIR PROCEDURES

~ DEDICATED REPAIR EQUIPMENT TOOLS AND MATERIALS (AVAILABLEON SITE)

~ REPAIRS FEASIBLE WITHIN APPLICABLE TIME REQUIREMENTS

TRIENNIALTEAM INSPECTION GENERAL GUIDANCE

~ TEAM MAKEUP

~ RX SYSTEMS ENGINEER

~ ELECTRICAL ENGINEER

~ FIRE P RQTECTION ENGINEER

~ FP REGULATORY BACKGROUND

~ INSPECTION PROCESS OVERVIEW

~ SIGNIFICANCE DETERMINATION PROCESS USES:

~ FOCUS TRIENNIAL INQUIRY

~ CHARACTERIZE FINDINGS e

TRIENNIALTEAM INSPECTION SPECIFIC GUIDANCE

~ SCOPE OF RESIDENT INSPECTOR MONTHLYTOURS (NON-DESIGN ORIENTED)

~ RESIDENT INSPECTOR USE OF THE SIGNIFICANCE DETERMINATION PROCESS

~ COMPENSATORY MEASURE ADEQUACY TO COMPENSATE FOR DEGRADED FUNCTIONS

~ INSPECTION PLAN DEVELOPMENT CONSIDERATIONS

0 Fire Protection Baseline Inspection Activities Resident Inspector and Triennial Team Preparation, Coordination, and Onsite Inspection Activities LEON WHITNEY PLANT SYSTEMS BRANCH NRR SEPTEMBER, 0 999

RESIDENT INSPECTOR PREPARATION o MONTHLYSELECTION OF 2-4 PLANT AREAS BASED ON:

~ PLANT-SPECIFIC RISK INFORMATION MATRIX, OR

~ 6 ENE Rl C RIM2 DOCUMENT COORDINATION

~ ANNUAL FIRE BRIGADE DRILL OBSERVATION IN A'PLANT AREA IMPORTANT TO SAFETY

~ REVIEW DRILL SCENARIO

~ REVIEW FIRE PROTECTION PRE-PLAN STRATEGY FOR SUBJECT PLANT AREA

RESIDENT INSPECTOR INSPECTION ACTIVITIES

~ MATERIALCONDITION AND OPERATIONAL STATUS OF:

~ TRANSIENT COMBUSTIBLES AND IGNITION SOURCES

~ FIRE DETECTION AND FIRE SUPPRESSION EQUIPMENT

~ FIRE BARRIERS

~ NO DESIGN FOCUS

~ SIGNIFICANCE DETERMINATION PROCESS APPLIED IF FIRE PROTECTION DEFENSE IN DEPTH IS IN QUESTION

TRIENNIALTEAM PREPARATION

~ INSPECTION NOTIFICATION LETTER TO LICENSEE ADDRESSES:

~ SCOPE OF INSPECTION

~ INFORMATION GATHERING, ACCESS TRAINING, 8 TEAM MEMBER BADGING VISIT

~ DOCUMENTATION NEEDS

~ PRE-INSP. CONFERENCE CALL ON ADMINSTRATIVE ITEMS AND ONSITE ACTIVITIES AND SCHEDULES

TRIENNIALTEAM PREPARATION

~ EXAMPLES OF SUPPORTING INFORMATION TO BE REVIEWED AND/OR OBTAINED:

~ CURRENT FP PROGRAM, FIRE HAZARDS ANALYSIS, AND FP IMPLEMENTING PROCEDURES DOCUMENTS

~ FIRE BRIGADE PRE-PLANS

~ REDUNDANT/ALTERNATIVE SAFE SHUTDOWN SYSTEMS SELECTIONS AND SEPARATION ANALYSES

TRIENNIALTEAM PREPARATION P

(SUPPORT INFO. CONTINUED)

~ FLOW DIAGRAMS AND PLANT EQUIPMENT LAYOUT DRAWINGS

~ FIRE ROOM, AREA AND ZONE BOUNDARY LAYOUTS

~ AUTOMATICSUPPRESSION AND DETECTION, AND MANUALSUPPRESSION EQUIPMENT LOCATIONS

\

~ EMERGENCY LIGHTING LOCATIONS

TRIENNIALTEAM PREPARATION (SUPPORT INFO. CONTINUED)

~ ASSOCIATED CIRCUITS ANALYSES

~ SELECTED FIRE PROTECTION FEATURE MAINTENANCEAND SURVEILLANCE PROC.

(INCLUDING TESTING OF FUSE BREAKER COORD.)

~ SELECTED FP AND PFSS/D RELATED DESIGN CHANGE PACKAGES

TRIENNIALTEAM PREPARATION (SUPPORT INFO. CONTINUED)

~ RCP OIL COLLECTION SYSTEM LAYOUTAND POTENTIAL RCP OIL SYSTEM LEAKAGE POINTS

~ SERS AND 50.59/GL 86-10 REVIEWS AND OTHER DOCUMENTS FORMING LICENSING BASIS FOR PFSS/D CONFIGURATION

~ CONFIGURATION CONTROL PROCEDURES AND INSTRUCTIONS

TRIENNIALTEAM PREPARATION (SUPPORT INFO. CONTINUED)

~ LISTING OF APPLICABLE CODES AND STANDARDS (E.G., NFPA CODES), AND CODE DEVIATION EVALS.

~ FP AND PFSS/D QA AUDITS AND QA SURVEILLANCES OF FP ACTIVITIES

~ OPEN AND CLOSED FP CONDITION REPORTS

~ COLD SHUTDOWN REPAIR ANALYSES, PROCEDURES, EQUIPMENT AND MATERIALS

TRIENNIALTEAM PREPARATION

~ REG. SENIOR RISK ANALYST (SRA) RISK INSIGHTS REPORT:

~ FIRE AREA RISK RANKINGS

~ CONDITIONALCORE DAMAGE P ROBAB ILITIES

~ TRANSIENT SEQUENCES

~ TEAM MEMBER TECHNICAL INPUT TO TEAM LEADER

~ TEAM LDR INSPECTION PLAN DEVELOPMENT (CONSIDERING PAST INSPECTION FINDINGS)

~ TEAM LDR SELECTION OF 3-5 AREAS IMPORTANT TO RISK

TRIENNIALTEAM COORDINATION AND ONSITE ACTIVITIES

~ PRE-INSPECTION CONFERENCE CALL:

~ ADMINISTRATIVEITEMS. FOR EXAMPLE, INSPECTION WEEK AVAILABILITY OF:

~ LICENSEE SUPPORT STAFF, COUNTERPARTS, TECHNICAL SPECIALISTS

~ INSPECTION SUPPORT DOCUMENTATION

~ TEAM MEETING ROOM

TRIENNIALTEAM COORDINATION AND ONSITE ACTIVITIES PRE-INSPECTION CONFERENCE CALL (CONTINUED)

~ FINALIZE PLANS AND SCHEDULES FOR:

~ OBSERVATION OF FIRE BRIGADE DRILL ALTERNATIVESAFE S/D PLANT WALKTHROUGH (SIMULATOR S/D NOT NORMALLY OBSERVED)

~ COLD SHUTDOWN REPAIR WALKTHROUGH

ATTACHHENT 71111. 05 INSPECTABLE AREA: Fire Protection CORNERSTONES: Initiating Events (10). Mitigating Systems (90) r INSPECTION BASES: Fire is gener ally a significant contributor to reactor plant risk. In many cases. the risk posed by fires is comparable to or exceeds the risk from internal events. The fire protection program shall extend the concept 'of defense in depth (DID) to fire protection in plant areas important to safety by (1) preventing fires from starting. (2) rapidly detecting.. controlling, and extinguishing those fires that do occur, and (3) providing protection for structures. systems, and components important to safety so that a fire that is not promptly extinguished by fire suppression activities will not prevent the safe shutdown of the reactor plant. If DID is not maintained by an adequately implemented fire protection program. overall plant risk can increase.

This inspectable area verifies aspects of the Initiating Events and Hitigating Systems cornerstones for which there are no performance indicators to measure licensee performance.

LEVEL OF EFFORT: For one hour a month, the resident inspector will tour from two to four plant areas important to reactor safety to observe conditions related to: (1) licensee control of'ransient combustibles and ignition sources; (2) the .material condition, operational status, and operational lineup of'ire protection systems. equipment and features; and (3) the fire barriers used to prevent fire damage or fire propagation. Once a year the resident inspector will observe a plant fire drill.

In addition, for one week every 3 years. in from three to five selected plant areas. an inspection team consisting of a fire protection engineer, a Issue Date: 09/13/99 DRAFT 71111. 05 DRAFT

reactor systems engineer, and an electr ical engineer will conduct a risk-informed. onsite inspection of the DID elements used to mitigate the consequences of' A'e. with emphasis on the A'e protection features provi'ded for maintaining at least one safe shutdown success path free of tire damage.

71111.05-01 INSPECTION OBJECTIVE The inspection objective is to assess whether. the licensee has implemented a fire protection program that adequately controls combustibles and ignition sources within the plant. provides adequate fire detection and suppression capability, and ensures that procedures, equipment, fire barriers, and systems exist so that the capability to safely shut down the plant is ensured.

711'11.05-02 INSPECTION REQUIREMENTS 02.01 Monthl Routine Ins ection.'or one hour each month, the resident inspector will tour from two to four plant areas important to safety to assess the material condition of'lant fire protection systems and features, their operational status, and the operational lineup of Are protection systems or equipment. The tour should concentrate on the material condition of fire detection and suppression systems and equipment, and on passive fire protection

~

features. For the areas selected, as applicable to the area of'oncern. conduct

~ ~

the following lines of inspection inquiry:

a. Control of'ransient Combustibles and I nition Sources

1. Observe if any transient combustible materials are located in the area. If transient combustible materials are observed. verify that they are being controlled in accordance with the licensee's administrative control procedures.

Observe if any welding or cutting (hot work) is being performed in the area. Verify that hot work is .being done in accordance with the licensee's administrative control procedures.

b. Fire Detection S stems. Verify that the fire detectors installed in the room are located near or on the ceiling. Observe the physical condition of the A'e detection devices and note any that show physical damage.

Determine from licensee administrative systems the known operational status of the system, and verify that any observed conditions do not affect the operational capability of the system.

71111.05 DRAFT Issue Date: 09/13/99 DRAFT

c. Fire Su ression S stems S rinkler Fire Su ression S stems. Observe that sprinkler heads are located near the ceiling and under major overhead equipment obstructions (e.g., ventilation ducts). Observe and verify that the water supply control valves to the system are open and that the fire water supply and pumping capability is operable and capable of supplying the water supply demand of the system. Observe and note any material conditions that may affect performance of the system, such as mechanical damage, painted sprinkler heads. or corrosion, etc.

Gaseous Su ression S stems. Observe that the gaseous suppression system (e.g. Halon or C02) nozzles are located near the ceiling and are not obstructed or blocked by plant equipment. Observe and verify that the suppression agent charge pressure is within the

. normal band. extinguishing agent supply valves are open. and that the system is in the automatic mode. Observe and verify that the dampers/doors will close automatically (or their closure is otherwise assured) upon actuation of the gaseous system. Observe and verify that the room penetration seals are sealed and in good condition. Observe and note any material conditions that may ai'feet performance of the system. such as mechanical damage, corrosion, damage to doors or dampers, open penetrations. or nozzles blocked by plant equipment.

d. 'Manual Fire fi htin E ui ment and Ca abilit Fire Extin uishers. Ensure that adequate numbers and types of portable Are extinguishes are provided at designated places in or near the area being inspected. and that access to the 're extinguishers is unobstructed by plant equipment or other work related, activities. Observe and verify that the general condition of fire extinguishes is satisfactory (e.g., pressure gauge reads in the acceptable range, nozzles are clear and unobstructed, charge test records indicate testing within the normal periodicity).

Hose Stations and Stand i es.'bserve and ver ify that a hose station can provide coverage for the area being inspected (maximum hose length 100 feet hand an electrically safe fog nozzle). Observe and verify that the water supply control valves to the standpipe system are open and that the Are water supply and pumping capability is oper able and capable of supplying the water flow and pressure demand. Ensure that access to the hose stations is

.unobstructed by plant equipment or work-related activities. Observe and verify that the general condition of hose stations is Issue Date: 09/13/99 DRAFT 71111.05 DRAFT

satisfactory (e.g., no holes in or chafing of the hose, nozzle not mechanically damaged and not obstructed, valve hand wheels in p lace).

e. 'assive Fire Protection Electrical Racewa Fire Barrier S stems. Observe the material condition of electrical raceway fire barrier systems (e.g. cable tray fire wraps) and determine if there are any cracks. gouges. or holes in the barrier material, that there are no gaps in the material at joints or seams, and that banding, wire. tie, and other fastener pattern and spacing appears appropriate. Where the fire material barrier is a wrap, or blanket-type material, observe that the

.has no tear s . rips, or holes in any of the visible layered material, that there are no gaps in the material at joint or seam locations, and that banding spacing appears appropriate. If plant modifications have recently been conducted, establish that A'e barriers removed as interfe'rence have been restored.

Fire Doors. Observe the material condition of the Are door in the area being inspected. Verify that selected fire doors close without gapping, and that the door latching hardware functions securely.

a Ventilation S stem Fire Dam ers. Observe the condition of the fire dampers in the areas being inspected. Ensure fusible link fire dampers are not prematurely shut or obstructed.

Structural Steel Fire Proofin . Observe the material condition of the structural steel fire-prooA'ng (fibrous or concrete encapsulation) within the areas being inspected. Verify that this material is installed and that the structural steel is uniformly covered.

Fire Barrier Electrical Penetration Seals. Tour plant areas being inspected and observe electrical and piping penetrations. Observe whether any seals are missing from locations in which they appear to be needed to complete a Are barrier. and determine that seals appear to be properly installed and in good condition. Verify that Are resistive material has been used to fill the opening/penetration' Com ensator Measures. Verify that adequate compensatory measures are put in place by the licensee for degraded or inoperable fire protection equipment, systems or features (e.g.. detection and suppression systems ,

and equipment, passive fire barrier features, or safe shutdown functions or capabilities).

71111. 05 DRAFT 4,- Issue Date: 09/13/99 DRAFT

02.02 Annual Routine Ins ection. During the annual obser'vation of a fire brigade drill in a plant area important to safety; the resident inspector should observe that:

Protective clothing/turnout gear is properly donned.

b. Self-contained breather apparatus (SCBA) equipment is properly worn and used.

C. Fire hose lines are capable of reaching all necessary fire hazard locations, that the lines are laid out without flow constrictions, the hose is simulated being charged with water, .and the nozzle is pattern (flow stream) tested prior to entering the fire area of concern.

d. The fire area of concern is entered in a controlled manner (e.g., fire brigade members stay low to the floor and feel the door for heat prior to entry into the fire area of concern).

Sufficient fire fighting equipment is brought to the scene by the fire brigade to properly perform their firefighting duties.

P The fire brigade leader's fire Aghting directions are thorough, clear, and effective.

Radio communications with the plant operators and between fire brigade members are efficient and effective.

Nembers of the fire brigade check for fire victims and propagation into .

other plant areas.

Effective smoke removal operations were simulated.

The Are Aghting pre-plan strategies were utilized.

k. The licensee pre-planned the drill scenario was followed. and that the drill objectives acceptance criteria were met.

02.03 Triennial Ins ection. Every three years in from three to five selected plant areas an inspection team will conduct a one-week inspection of the licensee's fire prote'ction program with emphasis on post-fire safe shutdown capability and the fire protection features provided for ensuring that at least one post-Are safe shutdown success path is maintained free of fire damage.

a, Iris ection Pre aration Issue Date: 09/13/99 DRAFT 71111. 05 DRAFT

4 Prior to the inspection information gathering trip...the regional senior reactor analyst (SRA) will provide the team leader with a summary of'lant specific fire risk insights (e.g.. fire risk ranking of the rooms/plant fire areas, conditional core damage probabilities (CCDPs) for those rooms and areas. and transient sequences for. these rooms). After considering the focus of past fire protection and post-tire safe shutdown inspections, the team leader will select three to five areas important to risk for team attention.

2. The inspection team leader will manage and coordinate a two or three day. information gathering site visit accompanied by the team members. The fire protection and post-fire safe shutdown information gath'ered by the team will center on the three to five areas selected by the team leader. During the reactor site visit all team members will receive site specific site access training and will be processed f'r unescorted site access.

After the information gathering site visit. the team leader will use the SRA developed fire risk insights, as well as technical input from the other team members, to develop an inspection plan addressing (for the selected three to five plant areas, rooms or zones) post-fire safe shutdown capability and the fire protection features provided f'r maintaining one train of this capability free of fire damage.

b. Ins ection Conduct. For the plant areas selected f'r review, conduct the following inspectiori,efforts:

S stems Re uired to Achieve and Maintain Post-tire Safe Shutdown Verify that the licensee's shutdown methodology has properly identified the components and systems necessary to achieve and maintain safe shutdown conditions. This requires verifying the following:

(a) The reactivity control function is capable of achieving and

maintaining cold shutdown reactivity conditions.

(b) .The reactor coolant makeup function is capable of maintaining the reactor coolant level above the top of the core for boiling water reactors (BMRs) or within the level indication in the pressurizer (or .solid plant) for pressurized water reactors (PMRs).

71111.05 DRAFT Issue Date: 09/13/99 DRAFT

4 (c) The reactor heat removal function is capable of achieving and maintaining decay heat removal.

(d) The process monitoring equipment provides direct readings of the process variables for reactivity control. coolant makeup, and decay heat removal -functions [note: source range neutron indication is not necessarily required, and an alternative method of reactivity measurement can be providedj.

(e) The support system functions are capable of providing the process cooling, lubrication, and other services necessary to permit extended operation of'he equipment used to accomplish sate shutdown functions.

2. Fire Protection of Safe Shutdown Ca abilit h

For the plant areas selected, evaluate the separation of systems necessary to achieve safe shutdown, and verify that fire protection features are in place to satisfy the separation and design requirements of Section III.G of Appendix R (or, for reactor plants reviewed under the Standard Review Plan,

, license specific requirements).

Verify that the fire detectors and automatic fire suppression systems, associated with 1-hour fire barriers required by Section III.G.2 of Appendix R (or, for reactor plants reviewed under the Standard Review Plan, license specific requirements), have been adequately installed. Review licensee evaluations which confirm that the installed automatic suppression systems would adequately suppress fires associated with the hazards of each selected area.

For the plant areas . selected, verify that redundant trains of systems required for hot shutdown located in the same tire area are not subject to damage from fire suppression activities or from the rupture or inadvertent operation of'ire suppression systems.

Determine each of the following:'a)

Whether a fire in a single location could, indirectly, thr ough the production of smoke, heat, or hot gases, cause activation of potentially damaging f'ire suppression for all redundant trains.

(b) Whether a fire in. a single location (or inadvertent actuation or rupture ot a fire suppression system) could, through local fire suppression activity, indirectly cause damage to all Issue Date: 09/13/99 DRAFT - 7- 71111. 05 DRAFT

redundant trains (e.g., sprinkler-caused flooding of other than the locally affected train), and (c) Whether, in response to a fire in a single location. the utilization of manually controlled fire suppression systems could cause damage to all redundant trains.

For the- plant areas selected, verify the adequacy of the design of fire area boundaries (i.e., able to contain the fire hazards of the area). raceway fir e barriers, equipment fire barriers, and fixed fire detection and suppression systems.

Address operator recovery actions (e.g., smoke removal, dewatering of spaces. controlled re-energization. and return to service of equipment in fire-affected areas) for fires in each plant area.

The observation of a fire brigade drill for a simulated fire in a plant area important to risk may be necessary to assess the effectiveness of manual fire fighting capability.

C Verify that adequate compensatory measures are put in place by the licensee for degraded or inoperable f'ire protection equipment.

systems or features (e.g., detection and suppression systems and equipment, passive fir~ barrier features, or safe shutdown functions or capabilities)

3. Post-fire Safe Shutdown Circuit Anal sis Verify that safety-related and non-safety-related cables for equipment in selected fire areas have been identified by the licensee and analyzed to show that they would not prevent safe shutdown because of hot shorts, open circuits, or shorts to ground.

Inspect the licensee's electrical systems and electrical circuit analyses with respect to the following:

(a) Common Power Su 1 /Bus Concern. On a sample basis, review the electrical distribution system and verify protective device coordination and that the licensee has analyzed the systems f'r high impedance fault conditions.

(b) Common Enclosure Concern. On a sample basis. review electrical fault protection from nonessential circuits routed in common enclosures (e.g.. fire wrapped electrical raceways) with required safe shutdown circuits.

71111. 05 DRAFT Issue Date: 09/13/99 DRAFT

(c) S urious Si nal Concern. On a sample basis review fire-induced hot shorts. shorts to ground, and open circuits and their potential effects on post-fire safe shutdown capability.

(d) Fuse/Breaker Coordination. On a sample basis. verify that circuit breaker coordination and fuse protection have been.

analyzed .and provided.

4. Alternative Shutdown Ca abilit N

I Verify the adequacy of the design and implementation of the licensee's alternative shutdown capability for selected plant areas by reviewing the licensee's alternative shutdown methodology and determining the identified components and systems necessary to achieve and maintain safe shutdown conditions. Establish that these components and systems can meet the following functional requirements:

(a) The reactivity control function is capable of achieving, monitoring, and maintaining cold shutdown reactivity conditions.

(b) The reactor coolant makeup function is capable of maintaining the reactor coolant level above the top of the core for BWRs, or is within the level indication in the pressurizer (or solid plant) for PWRs.

(c) The reactor heat removal function is capable of achieving and maintaining decay heat removal.

(d) The process monitoring equipment provide direct readings of the process variables for reactivity control, coolant makeup and decay heat removal functions [note: source range neutron indication is not necessarily required, and an alternative method of reactivity measurement can be providedj, and (e) The support system functions are capable of providing the process cooling, lubrication, and other services necessary to permit extended operation of the equipment used to provide safe shutdown functions.

Verify that hot and cold shutdown from outside the control room can be achieved and maintained with off-site power available or not available.

Is'sue Date: 09/13/99 DRAFT 71111. 05 DRAFT

Verify that the transfer of control from the control room to the alternative location has been demonstrated to not be affected by fire-induced circuit faults (e.g. by the provision of separate fuses and power supplies for alternative shutdown control circuits).

0 erational Im lementation of Alternative Shutdown Ca abilit Verify that the training program i'or licensed and non-licensed personnel has been expanded to include alternative or dedicated safe shutdown capability.

Verify that personnel required to achieve and maintain the plant in hot shutdown following a fire using the alternative shutdown system can be provided from normal onsite staff, exclusive of the fire brigade.

Verify that adequate procedures for use of the alternative shutdown system exist. Verify that the operators can reasonably be expected to perform the procedures within applicable shutdown time requirements. Ensure that adequate communications are available for the personnel performing alternative or dedicated safe shutdown.

Verify the implementation and human factors adequacy of the alternative shutdown procedures by "walking through" of the procedural steps.

Verify that the licensee has incorporated the operability of alternative shutdown transfer and control functions into the plant technical specifications.

Verify that the licensee periodically performs operability testing of the alternative shutdown instrumentation and transfer and control functions. In addition, verify that if'he licensee imposes the appropriate compensatory measures during periods in which alternative shutdown capability may be declared inoperable.

Communications Ver ify through observation of licensee conducted communication tests that portable radio communications and/or fixed emergency communications systems are available. operable, and adequate for the performance of alternative safe shutdown functions. Assess the ability of the communication systems to support the operators in the conduct and coordination of their required actions (e.g., consider ambient noise levels. clarity of reception, reliability and coverage patterns).

71111. 05 DRAFT Issue Date: 09/13/99 DRAFT

7. Emer enc 'Li htin Review emergency lighting provided for alternative safe shutdown along access routes and egress routes, and at control stations, plant parameter monitoring locations, and manual operating stations:

(a) If emergency lights are powered from a central battery verify that the distribution system contains or'atteries, protective devices so that a fire in the area will not cause loss of emergency lighting in any unaffected area needed for safe shutdown operations.

(b) Review 'the manufacturer's information to verify that battery power supplies 'are rated with at least an 8-hour capacity.

(c) Determine if the operability te'sting and maintenance of the lighting .units follow the manufacturer's recommendations.

(d) By asking the licensee to perform an emergency lighting test for selected plant areas. verify that the lamps, are properly aimed.

(e) Verify that emergency lighting unit batteries are being properly maintained (observe the unit's lamp or meter charge rate indication, and specific gravity indication).

(f) Verify that sufficient illumination is provided to permit access for the monitoring of safe shutdown indications and/or .

the proper operation of safe shutdown equipment. In coordination with the licensee, observe a normal station lighting blackout condition test in selected plant locations (e.g.. remote shutdown panel, switchgear room, diesel generator area). Determine if'llumination is adequate to perform required shutdown actions.

(g) Review the preventive maintenance surveillance procedure used for periodic checks of the emergency lights and verify that the maintenance frequencies and procedures are as specified by the manufacturer. Verify that the lighting units are routinely tested, and the testing criteria includes a "as-tound" manufacturers recommended discharge test.

8. Cold Shutdown Re airs Verify that the licensee has dedicated repair procedures, equipment, and materials to accomplish repairs of damaged components required Issue Date: 09/13/99 DRAFT 71111.05 DRAFT

for cold shutdown. that these components can be made operable, and that cold shutdown can be achieved within time frames specified by Appendix R to 10 CFR Part 50 (or, for reactor plants reviewed under the Standard Review Plan, license specific requirem'ents). Verify that the repair equipment, components, tools, and materials (e.g.,

pre-cut cable connectors with prepared attachment lugs) are available on site.

71111.05-03 INSPECTION GUIDANCE General Guidance Resident Ins ector Routine Monthl Ins ection. The main focus of the resident inspector 's activities is on the material condition and operational status of fire detection and suppression systems and equipment. and fire bar riers used to prevent fire damage or A'e propagation. The two to four plant areas to be inspected will be selected on the basis of the plant-specific risk information matrix, or the generic RIN2 document for the subject reactor plant.

Triennial Ins ection

~ob'ective. The one week, onsite. triennial inspection is primarily a risk-informed look at the mitigation elements of'ire protection defense in depth (DID) (i.e., detection, suppression, and confinement of fires through passive barriers, and the fire protection features and procedures which establish the licensee's ability to achieve and maintain post-fire safe shutdown conditions during and after a fire). The triennial inspection is uniquely that portion of the baseline inspection program that focuses on the design of reactor plant fire protection and post-fire safe shutdown systems, features'nd procedures. The inspection team leader will manage and coordinate the conduct of'n inspection emphasizing post-fire safe shutdown. The team will use plant-specific risk, event. and: technical information (including the results of licensee self-assessments) to confirm that at least one train of'afe shutdown equipment (capable of providing reactivity control, reactor coolant makeup, reactor heat removal. and process monitoring and support functions) is free of Are damage.

Post-fire Safe Shutdown Ca abilit Ins ection To ics. The confirmation of reactor plant post-fire safe shutdown capability includes (1) the identiAcation of safe shutdown'systems required to achieve the performance goals for the reactor plant's necessary shutdown functions.. (2) identification and design adequacy of'hysical separation (e.g., Fire Barriers) and suppression schemes used by the licensee to protect redundant cables or components (e.g., 3-hour barriers, 1-hour barrier/detection/suppression combinations. distance, exemption approved unique separation and suppression configurations); (3) review of the rating and physical condition of fire area boundaries to ensure their adequacy 71111. 05 DRAFT - 12- Issue Date: 09/13/99 DRAFT

.to contain the Are hazards within each fire area; (4) analysis of potential fire damage to power, control and indication cables for required systems so as to establish their continued ability to perform their iritended functions (5) review of electrical control transfer mechanisms f'r alternative safe shutdown capabilities at remote shutdown panels and/or emergency control stations (typically for postulated main control room and cable spreading room fires); (

6) review of'lternative or dedicated post-tire safe shutdown procedures, equipment access. communications and manual actions: (7) review of licensee circuit analyses for required and associated circuits of concern that could interfere with post-fire safe shutdown; and (8) review of cold shutdown equipment repair procedures. tools, and materials.

Ins ection A roach. The inspection of post-Are safe shutdown capability and its associated fire protection features can be either plant area-based or safe shutdown system-based, depending on the structure of the licensee's analysis.

Y Ins ection Team Makeu and Res onsibilities. The team assigned to conduct the multi-disciplinary triennial fire protection inspection will be comprised of a fire protection engineer, an electrical engineer, and a reactor systems engineer.

1. Reactor S stems En ineer SE . The reactor systems engineer will assess the capability of'eactor and balance-of-plant systems, equipment, operating personnel. and procedures to achieve and maintain post-Are safe shutdown and minimize the release of radioactivity to the

-environment in the event of A'e. He will be knowledgeable regarding integrated plant operations. maintenance, testing, surveillance and quality assurance, reactor normal 'and off-normal operating procedur es, and BMR and/or PMR nuclear and balance-of-plant systems design.

2. Electrical En ineer EE . The EE will identify electrical separation requirements for redundant train power, control, and instrumentation cables. He will verify that the licensee has adequately demonstrated that fire-induced circuit failures(hot shorts, shorts to ground, and open circuits) will not prevent safe shutdown operation.'e will review alternative shutdown panel electrical isolation design to establish the panels'lectrical independence from postulated A'e areas. He will also review required and associated circuits of concern for the elimination of fire-induced faults that can cause spurious signals which could interfere with post-fire safe shutdown, and in regard to common enclosure concerns and common power supply concerns. He will be knowledgeable regarding reactor plant electrical and instrumentation and control (I&C) design and will be familiar with industry ampacity derating standards

3. Fire Protection En ineer FPE . The FPE will review fire protection systems, features and procedures. The FPE will work with other team members in determining the effectiveness of the fire ba'rriers and systems Issue Date: 09/13/99 DRAFT 71111. 05 DRAFT

that establish the reactor plant's post-A'e safe shutdown conf'iguration and maintain it free of fire damage. He will determine whether suitable fire protection features (suppression, separation distance, fire barriers, etc.) are provided for the separation of equipment and cables required to ensure plant safety. Ke will possess a fire protection degree or equivalent experience, and will be knowledgeable regarding.

reactor plant fire protection systems and features.

Re ulator Re uirements and Licensin Bases. The regulatory requirements and licensing bases against which post-fire saf'e shutdown capability is assessed are as follows:

Plants licensed before Januar 1 1979. Effective February 17, 1981. the NRC amended its regulations by adding Section 50.48 and Appendix R to 10 CFR Part 50 to require certain provisions f'r fire protection in nuclear power plants licensed to operate before January 1, 1979. This action was taken to resolve certain contested generic issues in fire protection safety evaluation reports (SERs) and to require all applicable licensees to upgr ade their plants to a level of fire protection equivalent to the technical requirements in Sections III.G, J. L. and 0 of 10 CFR Part 50.

Appendix R. Licensees were required to meet the separation requirements of Section III.G.2. the alternative or dedicated shutdown capability requirements of Sections III.G.3 and III.L. or to request an exemption in accordance with 10 CFR 50.48. Alternative or dedicated safe shutdown capabilities were required to be submitted to the Office of Nuclear Reactor Regulation (NRR) f'r review. NRR approvals are documented in SERs.

Plants licensed after Januar 1 1979: These plants are subject to requirements similar to those in 10 CFR part 50, Appendix R, as specified in the conditions of their facility operating license, commitments made to the NRC, or deviations. granted by the NRC. These reactor plants licensed after January'. 1979. are subject to 10 CFR 50.48 (a) and (e) only.

The fire hazards analysis (FHA) ("Fire Protection Revieq, Fire Protection Evaluation" ) document of the reactor plants licensed after January 1, 1979, may have been reviewed under Appendix A to Branch Technical Position APCSB 9.5-1, "Guidelines for Fire Protection for Nuclear power Plants Docketed Prior to July 1. 1976." of August 23. 1976 (in which case, the licensee. conducted an Appendix R comparison and justified final safety analysis report (FSAR) or FHA differences from the specific provisions of Appendix R). It is possible, also that licensee submittals for plants licensed after January 1, 1979, were reviewed under the Standard Review Plant. NUREG-0800, and Branch Technical Position (BTP)

CMEB 9.5-1 (former ly BTP ASB 9.5-1), "Guidelines for Fire Protection for 71111. 05 DRAFT Issue Date: 09/13/99 DRAFT

Nuclear Power Plants, Rev. 2 (July 1981) (in which case, licensee submittals were reviewed accor ding to requirements that closely paralleled the provisions of Appendix R).

4 The actual fire protection requirements applicable to a given reactor plant licensed after January 1, 1979, arise from the specific license.

conditions in the facilit'y operating license. These license conditions possibly refer to SERs and their supplements. Section 9.5 of'uch an SER delineates which licensee submittals were reviewed (e.g., a fire hazards analysis would be such a submittal). The plant configurations and procedures described in these submittals are "requirements of the license."

Ins ection Process Licensee Notification Letter. The licensee should be notified of the triennial inspection in writing at least three months in advance of the

'nsite week. The letter should discuss the scope of the inspection, request an information-gathering visit to the licensee reactor site/engineering offices, discuss documentation and licensee personnel availability needs during the onsite inspection week, and request a pre-inspection conference call to discuss administrative matters and finalize inspection activity plans and schedules.

2. Information- atherin Site Visit. The inspection team leader will manage and coordinate a two to three day information gathering site visit accompanied by the team members. The purposes of the information gathering site visit are to (1) gather site-specific information important to inspection planning. (2) conduct initial discussions with licensee representatives regarding administrative items and inspection activity plans and schedules, and (3) have the team members receive site specific access training and badging for unescorted site access. In advance of the information-gathering site visit, and in order for the onsite information exchange to be as effective as possible. the team leader should give the licensee a list of information and documents that

.may be needed to prepare for and conduct the inspection.

Information Re uired. The team members should gather sufficient information to become familiar with the following:

(a) The reactor plant's design. layout. and equipment configuration.

(b) The reactor plant's current post-fire safe shutdown licensing basis through review of 10 CFR 50.48. 10 CFR Part . 50 Appendix R (if applicable), NRC safety evaluation reports (SERs) on fire S

Issue Date: 09/13/99 DRAFT 71111. 05 DRAFT

protection, the plant's operating license. updated final safety analysis report (UFSAR), and approved exemptions or deviations.

(c) The licensee's strategy and methodology. and derivative procedures, for accomplishing post-fire safe shutdown conditions. Among the sources of information are the updated final safety analysis report.

(UFSAR), the latest version of the fire hazards analysis (FHA). the latest version of the post-fire safe shutdown analysis (SSA), tire protection/post-fire safe-shutdown related 10 CFR 50.59 and Generic Letter 86-10 review documentation and modification packages, plant drawings, emergency/abnormal operating procedures, and the results of licensee internal audits (e.g., self assessments and quality.

assurance (QA) audits in the fire protection and post-fire safe shutdown 'areas).

(d) The historical record of plant-specific fire protection issues through review of plant-specific documents such as previous NRC inspection results, internal audits performed by the reactor licensee (e.g., self-assessments and quality assurance audits).

corrective action system records, event notifications submitted in accordance with 10 CFR 50.72. and licensee event reports (LERs) submitted in accordance with 10 CFR 50.73.

(e) The safe shutdown systems and support systems credited by the licensee's, analysis for each fire area, room, or zone for accomplishing of the required shutdown functions (e.g., reactivity

'ontrol, reactor coolant makeup, reactor heat removal, and process monitoring and support functions) as necessary to comply with the safe shutdown requirements of 10 CFR 50.48(a) and plant-specific licensing requirements. The shutdown logic tor each area, room, or zone to be inspected must be thoroughly understood by the team members.

(f) The licensee's analytical approach for electrical circuits separation analyses, and the licensee's methodology for identification and resolution of associated circuits of concern.

The team's electrical review should include addressing the assumptions and boundary conditions used in the performance of the licensee's analyses.

4. Si nificance'etermination Process SDP . The inspection team may identify a finding or set of findings that call into question one or more elements of'efense in depth (DID) at the reactor plant. In order to make a determination of the significance of the finding(s), it may be necessary to evaluate them within the significance determination process in the referenced supplemental fire protect'ton functional inspection 71111. 05 DRAFT - 16- Issue Date: 09/13/99 DRAFT

procedure (the "Fire Protection Risk Significance Screening Hethodology" of'P XXXXX). The results of such significance evaluations can be used to help . the team leader to (1) develop the in-process information necessary to prioritize and focus further onsite inspection activities.

and (2) characterize the significance of triennial team inspection findings both during and after the site exit meeting with the licensee.

Specific Guidance 03.01 Ins ection Re uirement 02.01. The resident inspector should not attempt to address all plant areas each month. The monthly plant tour should focus on from two to four plant areas important to risk. The resident inspector should note transient combustibles and ignition sources (and compare these with the limits provided in licensee administrative procedures). The resident inspector should also note the material condition and operational status (rather than on the design) of fire detection and suppression systems. and fire barriers used to prevent i'ire damage or fire propagation.

The inspector may identify a finding or set of findings which call into question one or more elements of defense-in-depth at the reactor plant. In order to assess the degree of degradation of'he DID element(s). and make a determination of'he significance of the finding(s), it may be necessary to evaluate them within the significance determination process of the referenced supplemental fire protection functional inspection procedure (the "Fire Protection Risk Significance Screening Methodology" of IP XXXXX).

k 03.02 Ins ection Re uirement 02.0lf. Short term compensatory measures should be adequate to compensate for the degraded function or feature until appropriate corrective action can be taken.,

03.03 Ins ection Re uirement 02.03a3. The inspection plan issued by the team leader for the triennial inspectiqn should consider or contain the following:

1. Recognition of the limitations imposed by the short (1 week) duration triennial inspection site visit:

of'he

2. The adequacy of the time allocated for the conduct of'nspection efforts to gather information required for the application of the Fire Protection Risk Significance Screening Nethodology contained in the reference supplemental fire protection inspection procedure (see section 03.01 above).

3. Follow-up on results of recent fire protection inspections.

f'r if it is determined that corrective actions specific risk-important inspection findings from such inspections appeai to be deficient or inadequate.

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03.04 Ins ection Re uirement 02.03b2: Short term compensatory measures should be adequate to compensate for the degraded function or feature until appropriate corrective action can be taken.

71111.05-04 RESOURCE ESTIMATE This procedure is estimated at 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> per year for routine inspection and 108 hours0.00125 days <br />0.03 hours <br />1.785714e-4 weeks <br />4.1094e-5 months <br /> every 3 years for the triennial inspection .

71111.05-05 REFERENCES IP XXXXX. "Fire Protection Functional Inspection" Month gg, 1999.

71111. 05 DRAFT Issue Date: 09/13/99 DRAFT

ATTACHMENT ROUTINE INSPECTION GUIDANCE TABLE CORNERSTONE RISK PRIORITY EXAMPLES INITIATING EVENTS (10) Equipment or actions that could Transient combustibles (rags. wood.

cause or contribute to initiation ion exchange resin, lubricating of fires in plant areas important oil. or Anti-Cs) are not in areas to safety or near equipment where transient combustibles are required for safe shutdown. prohibited. Transient combustible amounts in other areas do not exceed administrative controls.

Ignition sources (welding.

grinding. brazing, flame cutting) have a fire watch. Planning includes precautions and additional fire prevention measures where these activities are near combustibles.

HITIGATING SYSTEHS (90) Functionality of fire barriers in Doors and dampers that prevent the plant areas important to safety. spread of fires to/or between plant areas important to safety remain in Functionality of detection systems place and are functional.

in plant area important to safety.

Electrical raceway fire barriers Functionality of automatic and penetration seals that protect suppression systems in plant areas the post-fire safe-shutdown train important to safety. are not damaged.

Fire brigade manual suppression Fire detection and alarm system is effectiveness. functional for plant areas important to safety.

Compensatory measures for degraded fire detection systems. fire Automatic suppression system suppression features. and barriers sprinklers are functional and their to fire propagation. sprinkler head patterns are not blocked by plant equipment.

Fire brigade performance indicates a prompt response with proper fire fighting techniques for the type of fire encountered.

Hanual fire suppression equipment is of the proper type and has been tested.

Degraded fire detection equipment.

suppression features and fire propagation barriers are adequately compensated for on reasonably short-term bases.

END Issue Date: 09/13/99 DRAFT 71111.05 DRAFT

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P O

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O I

~O 4~*~4 POLICY ISSUE (NEGATIVE CONSENT)

Ma 20 1999 SECY-99-140 FOR: The Commissioners FROM: William D. Travers Executive Director for Operations

SUBJECT:

RECOMMENDATION FOR REACTOR FIRE PROTECTION INSPECTIONS (WITS ITEM 199700021)

PURPOSE:

To provide information about fire risk and the significance of reactor fire protection; to inform the Commission about the completion of the fire protection functional inspection (FPFI) pilot program as directed in a staff requirements memorandum (SRM) of February 7, 1997; and to make recommendations on the appropriate types and frequencies of reactor fire protection inspections and whether they should be part of the new reactor oversight process as directed in an SRM of April 1, 1999.

BACKGROUND:.

Overview of Reactor Fire Risk Risk assessments have shown that fires in nuclear power plants can be risk significant. To ensure that its decisions and recommendations regarding the regulatory program for reactor .

fire protection are commensurate with its importance and risk sign Tiicance, the staff considers the underlying purpose of the regulatory requirements and, to the extent practicable, available CONTACTS:

Leon Whitney, SPLB/DSSA/NRR 301-415-3081 Steven West, SPLB/DSSA/NRR 301%1 5-1 220

The Commissioners fire risk data and insights. This paper presents an overview of the considerations applied by the staff.:

Fire Protection Functional Ins ection Pilot Pro ram The staff informed the Commission of its plans for implementing the fire protection functional inspection pilot program in SECY-96-267, "Fire Protection Functional Inspection Program,"

December 24, 1996. In an SRM of February 7, 1997, the Commission informed the staff that it did not object to the staff's plans for the pilot program. In SECY-98-187, "Interim Status Report

- Fire Protection Functional Inspection Program," August 3, 1998, the staff described the inspection results to date. The staff described its remaining work and the scope of its planned final report on the FPF I pilot program in SECY-99-040, "Second Interim Status Report - Fire Protection Functional Inspection Program," February 5, 1999. The staff briefed the Commission on February 9, 1999. In an SRM of April 1, 1999, the Commission directed the staff to include in its final report its recommendations on the appropriate types and frequencies of reactor fire protection inspections and whether they should be part of the new reactor oversight process. This paper presents the staff's final report on the FPFI pilot program and the staff's recomm'endations. This completes the staft's actions on WITS Item 199700021.

DISCUSSION:

Overview of Reactor Fire Risk Simply stated, the underlying purpose of the NRC fire protection regulation is to provide reasonable assurance that one means of achieving and maintaining safe-shutdown conditions will remain available during and after a fire. This is accomplished by applying the concept of fire protection defense in depth to reduce the likelihood of fires and to limit the extent of fire damage to the structures, systems, and components that would be used to achieve and.

maintain safe-shutdown in the event of a fire. The stated objectives of fire protection defense in depth are to prevent fires from'starting; to rapidly detect, control, and extinguish fires that do start; and to design and protect structures, systems, and components so that a fire that. is not promptly extinguished will not adversely affect safe-shutdown.

Fire is not treated as a design-basis accident, nor are fires postulated to occur simultaneously with non-fire-related failures in safety systems, plant accidents, or severe natural phenomena.

The regulation requires that only one train of equipment necessary to achieve hot shutdown be maintained free of fire damage. Unlike systems provided to mitigate design-basis'accidents, the fire protection regulation does not require redundant or diverse post-fire safe-shutdown methods. Nor does the regulation require that fire protection systems and feat'ures or the structures, systems, and components provided for achieving post-'fire safe-shutdown be safety related, protected against a single failure, or covered by technical specifications. Finally, the regulation does not require that the equipment provided to achieve cold'shutdown or to mitigate the consequences of design-basis accidents be maintained free of fire damage.

In 1989, Sandia National Laboratories issued the "Fire Risk Scoping Study." This study, which included a review of the fire probabilistic risk assessments (PRAs) for four plants, concluded that the most risk-sign Tiicant plant areas typically are main control rooms, cable spreading rooms, and switchgear rooms. According to this study, although plant modifications made in

qo The Commissioners response to Appendix R reduced, by a factor of 3 to 10, the core damage frequencies (CDFs) at the plants studied, fire can be an important contributor to'CDF even after regulatory criteria have been satisfied. The study suggests that without the existing regulatory requirements, fire risk could be higher than it is today. The study also suggests that improper implementation of the regulatory requirements and degradation of fire protection defense in depth could be risk significant. The study concluded, for example, that weaknesses in either manual fire fighting effectiveness or control systems interactions could raise the estimated fiire-iriduced CDF by an order of, magnitude.

Under the individual plant examination of external events (IPEEE) program, the licensees systematically assessed the fire risk for each operating reactor. Although most licensees have reported numerical fire CDF estimates, the staff has not validated the accuracy of such

, estimates. Because the licensees may have used simplifying assumptions and approximate procedures in the analyses, the quantified CDF estimates reported in the IPEEE submittals only serve as general indicators of plant fire risk. Nevertheless, the results of the IPEEE fire analyses provide important insights regarding reactor fire risk and confirm the results of the "Fire Risk Scoping Study." For example, the IPEEE results show that fire events are important contributors to the reported CDF for a majority of plants, ranging on the order of 1E-9/yr to 1EP/yr, with the majority of plants reporting a fire CDF in the range of 1E-6/yr to 1EQ/yr. The reported CDF contribution from fire events can in some cases approach (or even exceed) that from internal events. Licensees proposed or implemented procedural and/or hardware improvements in the fire area in response to the IPEEEs at more than half of the plants. (In most cases any risk reduction achieved by the improvements was not reported separately, but was included in the total CDF.) Overall; the IPEEEs have confirmed that main control rooms, cable spreading rooms, and switchgear rooms are usually the most risk-significant plant areas.

Although it is generally understood that fire events can be serious and risk significant, fire science is a relatively new field. NRC fire research efforts and fire risk assessments have yielded both useful tools and important results. However, a number of important questions remain regarding the assessment and assurance of nuclear fire safety. For example, there are still significant uncertainties in the ability to mechanistically predict the behavior of fires under the broad variety of conditions that are relevant to nuclear power plant safety.

Attachment 1, "Fire Risk Fact Sheet," gives additional information about reactor fire risk.

'ire Protection Functional Ins ection Pro ram Pre-FPFI Pilot Program Inspection Procedures Since Appendix R was issued in 1981, the staff published three reactor fire protection inspection procedures (IPs). These are IP 64100, "Postfire Safe Shutdown, Emergency Lighting and Oil Collection Capability at Operating and Near-Term Operating Reactor Facilities"; IP 64150, Triennial Post-Fire Safe Shutdown Capability Revenfication"; and IP 64704, "Fire Protection/Prevention Program."

During the 1980s, the staff performed a 1-week team inspection at each facility using IP 64100. These inspections consisted of an audit review of the plant fire protection features and post-fire safe-shutdown capability (hardware and procedures) against the licensee's

The Commissioners 4-commitments and the applicable regulatory requirements. The inspections did not include detailed inspections of the safe-shutdown analysis or of the design bases of fire detection systems, fire suppression systems, and fire barriers installed to protect safe-shutdown equipment. For example, the inspector would verify, on an audit basis, that sprinkler systems installed to protect safe-shutdown equipment in accordance with the regulatory requirements were installed in the appropriate fire areas, but did not verify that the systems met the code requirements. Since these one-time inspections, the staff has re-inspected fewer than 10 plants in accordance with IP 64150, a 1-week regional initiative team inspection. The regions conduct IP 64704, the routine (core) reactor fire protection inspection procedure, at each plant about once every 3 years. The objective of this IP, which is typically conducted by a regional inspector over about a 1-week period, is to inspect, on an audit basis, the overall adequacy of the licensee's fire protection program. The focus of IP 64704 is on such typical fire protection features as extinguishers, fire hose stations, and sprinkler systems. In response to the Thermo-Lag fire barrier and penetration seal issues, the staff recently added to this IP additional guidance for inspecting these features. However, IP 64704, does not address detailed design basis issues or post-fire safe-shutdown capability, nor does it thoroughly evaluate fire protection program configuration management.

Basis For and Scope of FPFI Pilot Program As documented In SECY-96-267, the FPFI program was based on the following staff commitments to the Commission: (1) to inspect the Thermo-Lag corrective actions at all plants, (2) to assess the NRC reactor fire protection program to determine if it had appropriately addressed all.fire safety issues, (3) to determine if licensees are maintaining compliance with NRC fire protection requirements, (4) to identify the strengths and weaknesses of the reactor fire protection program, (5) to reevaluate the scope of the reactor fire protection inspection program, and (6) to develop a coordinated approach for reactor fire protection and systems inspectio'ns.

The FPFI pilot program consisted of a new inspection procedure, four pilot inspections, a public workshop, and a final report. The staff conducted four pilot inspections, using fire risk insights to help focus the inspections on areas most important to safety. Three of the four pilot inspections were full-scope FPFls.'The fourth pilot inspection was a reduced-scope inspection of a licensee self-assessment. During the pilot program period, the staff also conducted major team inspections at Quad Cities and Clinton.. Attachment 2 is the "Fire Protection Functiorial Inspection Pilot Program Final Report." Among other things, the final report presents the detailed background of the FPFI pilot program, the program objectives and accomplishments, summaries of the FPFI findings and resulting enforcement actions, insights from the FPFI workshop, interactions with the Nuclear Energy Institute (NEI), and insights and lessons learned from the FPFI pilot program. For this paper and the final report, and for the the .

purpose of developing its plans for future reactor fire protection inspections, the staff considered the results of the four pilot FPFls and the results of the inspections at Quad Cities and Clinton.

The Commissioners Coordination With New Reactor Inspection and Oversight Program After it initiated the FPFI pilot program, the staff began to look for other ways to improve the NRC's reactor oversight process. To ensure that its plans for future reactor fire protection inspections would be consistent with the new reactor inspection and oversight program, the staff responsible for reactor fire protection and the FPFI program coordinated with the reactor oversight task groups and considered the concepts and objectives of the new program when it developed the plans presented below.

Significant Insights and Lessons Learned The staff developed the following insights and lessons learned from information it gathered during the FPFI pilot inspections, the team inspections of Quad Cities and Clinton, and the FPFI workshop.

As discussed during the Commission meeting on February 9, 1999, one of the results of the FPFI program was renewed industry attention to nuclear power plant fire safety. For example, in response to the FPFI pilot program a number of licensees conducted comprehensive self-assessments of their fire protection programs even though they had not been selected as pilot .

plants. In addition, in response to the FPFI pilot program, NEI is developing procedures to help licensees conduct self-assessments. (Additional information about this NEI effort is presented under "Ongoing Staff Work," below.) Finally, during NEI Fire Protection Information Forums and other forums the staff has received information from licensees about voluntary changes to fire protection programs and planned self-assessments in response to the lessons learned from the FPFI pilot program.

For each FPFI, a senior NRR risk analyst reviewed available IPEEE results and other sources of risk information. (The FPFI inspection procedure contains guidance for using risk information and insights to focus inspection activities.) The risk insights, which were used as input to the FPFI inspection plans, helped focus the FPFls on areas in which the potential fire risks were greater and helped the inspectors improve their understanding of the significance of inspection findings.

As noted in SECY-98-187, potentially risk significant FPFI findings related to the regulatory requirements and licensee commitments had not been and would not have been revealed using the current core fire protection inspection procedure (IP 64704). Similarly, licensee quality assurance audits of reactor fire protection programs had not uncovered many of the findings related to the regulatory requirements and licensee commitments that were revealed during the pilot FPFls..

As noted above, until the FPFI pilot program, the NRC reactor fire protection inspection procedures did not direct the staff to thoroughly inspect the design bases of fire detection systems, fire suppression systems, and fire barriers installed to protect safe-shutdown equipment in accordance with the regulatory requirements or the details of the post-fire safe-shutdown analyses performed by the licensees to demonstrate compliance with the regulatory requirements. The FPFls included findings associated with the designs of fire protection systems and with safe-shutdown capabilities, including actions taken by licensees to resolve Thermo-Lag fire barrier issues.

The Commissioners The FPFI pilot program inspection results and other indicators, such as licensee event reports, indicate deficiencies and weaknesses'in reactor fire protection and post-fire safe-shufdown programs. For example, all of the pilot plants had some fire brigade weaknesses. Other findings were, for example, inadequate safe-shutdown analyses, inadequate or incomplete circuit analyses (discussed separately, below), incomplete safe-shutdown procedures, and inadequate attention to fire protection program management. The FPFI pilot program results suggest that deficiencies could exist in one or more layers of fire protection defense in depth at any given, plant. As discussed above, because fires can be important contributors to CDF even after regulatory criteria have been satisfied, and because improper implementation of the regulatory requirements and degradation of fire protection defense in depth can be risk significant, it is important that the licensees maintain the reactor fire protection programs and that the staff monitor licensee performance in this area.

As noted above, the licensees for several of the FPFI pilot plants had conducted fire protection program self-assessments in advance of the FPFI. One of the pilot FPFls was a reduced-scope inspection of a licensee self-assessment. The self-assessments were based largely on the FPFI procedure and lessons learned by the licensees by observing previous pilot FPFls.

Overall, the licensee self-assessments were of good quality, were commensurate with the scope and depth of an FPFI, and reflected the strengths and weaknesses of the licensees'rograms fairly well.

Independent of the FPFI pilot program and before the staff began the pilot FPFls, the reactor industry raised questions about the adequacy of the existing staff guidance concerning fire-induced circuit failures and the consistency of staff interpretations of both the guidance and the underlying regulatory requirements. The staff and the industry are currently working to resolve these questions'and to develop new guidance, if needed, to resolve this issue. During several of the FPFls, the inspectors found compliance issues associated with the analyses that the licensees had performed to identify the circuits that require fire protection to suppoit post-fire safe-shutdown in accordance with the fire protection rule. These findings confirmed that plant-specific circuit analysis issues may exist because of differing staff and licensee interpretations of the existing guidance and regulatory requirements. However, during the FPFI pilot program, the staff did not identify any sign Tiicant new questions or issues concerning the existing fire protection regulatory requirements and guidance, or with the staff's licensing reviews of reactor fire protection programs. '(The staff notes that at the FPFI workshop, several participants expressed uncertainty as to what constitutes compliance with the fire protection requirements, indicating that the uncertainty stems from the complexity of reactor fire protection, questions about existing staff guidance, and changing staff expectations.

Circuit analysis was the only specific example offered. The ongoing staff and industry, activities to resolve the circuit analysis issue and the comprehensive regulatory guide for reactor fire protection that the staff is currently developing will address any uncertainties associated with the existing staff guidance and the regulatory requirements.)

Ongoing Staff Work Under the new reactor inspection and oversight program, fire protection (an area that includes fire protection features and post-fire safe-shutdown capability) falls within both the initiating events cornerstone and the mitigating systems cornerstone. Fire protection is not presently covered by performance indicators. Therefore, reactor fire protection has been identified as an

The Commissioners inspectable area under the net program. As described in SECY-99-007A, "Recommendations for Reactor Oversight Process Improvements (Follow-Up to SECY-99-007)," March 22, 1999,.

the staff has drafted procedures for baseline inspections of fire protection programs that it plans to test during the pilot program and then incorporate into the final program.

During a public meeting on March 25, 1999, the staff discussed with NEI and other interested stakeholders its proposed fire protection baseline procedures. By letter dated April 14, 1999, NEI submitted comments on the draft procedures and expressed concerns about including fire protection in the pilot program because it was only recently added and the licensees may not have time to fully plan for this aspect of the pilot program. In response to NEI's comments, the staff revised the draff procedures, as appropriate. During a public meeting on May 6, 1999, the staff met with NEI to discuss its letter of April 14, 1999, and gave its revised procedures to NEI. In response to NEI's concerns about i'ncluding fire protection in the pilot program, the staff and NEI discussed conducting any pilot fire protection inspections late enough in the program to allow for adequate planning and preparation by the staff and the pilot plant licensees. The staff indicated that it currently plans to conduct pilot fire protection inspections at three of the pilot plants. The specific plants have not yet been identified. The staff also indicated that it would take appropriate steps to ensure that the inspections were effective and efficient, including for example, meeting with the pilot plant licensees to discuss the fire protection inspection process and to address any unanswered questions. The staff and NEI agreed to meet again on May 24, 1999. The staff will continue to work with NEI and other stakeholders as appropriate.

At the time of the FPFI pilot inspections, a tool for systematically assessing the risk significance of fire protection inspection findings was not available. During the FPFI workshop, there was general consensus'that such a tool would be beneficial to both the staff and the industry. Subsequently, NRR's'Plant Systems Branch and Probabilistic Safety Assessment Branch, with assistance from the. Office of Nuclear Regulatory Research and the senior reactor

- analysts, developed a proposed method for assessing the potential fire risk significance of fire protection inspection findings. After it is completed, the Fire Protection Risk Sign Tiicance Screening Methodology (FPRSSM), which is described in Section 9 of the final report, could be

'sed by inspectors to focus on risk-signiTicant sets of inspection findings, while screening out findings that have minimal or no risk significance. The FPRSSM could also be used to

.evaluate inspection findings after the inspection. Inspection findings that the FPRSSM finds to be potentially risk significant (i.e., those that are not screened out as having minimal or no risk significance) could be subjected to a more refined evaluation to help establish the appropriate regulatory response to the findings. The FPRSSM could also be used by the licensees to assess deficiencies found during self-assessments.

Although the FPRSSM was not available during the FPFls or the resulting enforcement proceedings, as a final activity of the FPFI pilot program, the staff used the FPRSSM to assess a sample of the FPFI findings. The results of two FPRSSM assessments, one of potentially high risk significance and one of low risk significance, are summarized in Section 10 of the final report. The staff is working with the reactor oversight task groups to incorporate the FPRSSM into the Inspection Finding Risk Characterization Process described in SECY-99-007A. During a public meeting on March 25, 1999, and the NEI Fire Protection Information Forum on May 3, 1999, the staff discussed the proposed FPRSSM with NEI and other interested stakeholders. During a public meeting on May 6, 1999, the staff gave NEI a

The Commissioners copy of the draft FPRSSM forreview. The staff and NEI agreed to meetagain on May 24, 1999, to discuss any NEI comments and to work through some sample applications..

NEI provided input and feedback on the FPFI pilot program during the FPFI workshop, in

.correspondence, and during follow-up meetings with the staff. NEI favors more reliance on licensee self-assessments, with FPFls reserved for use when licensee performance approaches the "unacceptable performance," as defined by the new inspection and oversight program. An NEI issue task force is developing procedures (based on the FPFI procedure) to help licensees conduct self-assessments. It is the staffs understanding that NEI plans to phase in the self-assessment procedures, which will be available for licensees to use on a voluntary basis, beginning in summer 1999. While recognizing the difficulties of doing so, NEI has also formed an issues task force to develop performance indicators for reactor fire protection programs. NEI plans to complete small-scale performance indicator pilot trials at reactor sites by July 2000. During the public meetings on March 25 and May 6, 1999, the staff informed NEI that it would consider the results of NEI's efforts in the new reactor inspection and oversight program as they become available.

Conclusions Although the FPFI program involved a relatively small sample of plants, the results of the inspections, coupled with other indicators, such as licensee event reports, suggest that continued monitoring of licensee performance is needed to achieve confidence that risk-significant fire protection program deficiencies do not exist at any given plant. Inspection is a proven and appropriate means of both monitoring licensee performance and finding deficiencies. Therefore, on the basis of the insights and lessons learned from the FPFI pilot program, the importance of fire protection from the point of view of potential risk, past experience (e.g., such issues as Thermo-Lag and circuit analysis), the absence of

'perational fire protection and post-fire safe-shutdown performance indicators, and the existing regulatory requirements, the staff concludes that some level of NRC inspection of reactor fire protection programs is appropriate to maintain safety and increase public confidence. The staff also concludes that future NRC fire protection inspections should be consistent with the concepts and objectives of the new reactor inspection and oversight program and should be included within that program.

The staff believes that future fire protection inspections should be more comprehensive and risk informed than the current core inspections. For example, future inspections should address the existing regulatory requirements regarding post-fire safe-shutdown capability, which are not inspected under the current core inspection program, with emphasis on activities,.plant areas, and safe-shutdown configurations where the potential risks are greater.

In light of the renewed industry attention to reactor fire safety and NEI's plans for voluntary self-assessments, the staff also concludes that intense fire protection inspections, such as full-scope FPFls, are not warranted as a routine-type inspection, but should be available for use on an as-needed basis, such as when plant performance declines or to respond to a specific

'event or problem at a plant.

1 The staff believes that, over both the short and long terms, licensee self-assessment activity will increase. The number and significance of reactor fire protection program deficiencies should decrease in response to more frequent and more robust self-assessments and NRC

The Commissioners concludes that it would be monitoring through the baseline inspections. Therefore, the staff touche appropriate to consider licensee self-assessments during future NRC inspections, provided .

that the scope and depth of the self-assessments are equivalent scope and depth of the NRC inspections discussed below. In such cases, in addition to some independent verifications of fire protection program features, the NRC inspections would verify the accuracy of the licensees'ssessment processes, and would review the licensees'ffectiveness in maintaining the appropriate level of performance to assure safe operation, and in finding and resolving problems. In addition, taking into account the new reactor inspection and oversight program, if valid fire protection performance indicators are eventually developed; the staff will reassess the baseline fire protection inspection program {discussed in the following section) and consider changing the scope and frequency of the inspections, as appropriate. The staff's planned actions are presented below.

. RECOMMENDATION:

Unless otherwise directed by the Commission, the staff will:

include risk-informed baseline procedures for routine resident inspector walkdowns and for triennial fire protection team inspections within the new reactor inspection and oversight program to monitor licensee performance in the fire protection area. The staff previously described this approach in SECY-99-007A.

2. Structure the triennial baseline inspection procedure to emphasize its modular nature so that it could be used, for example, to independently inspect Thermo-Lag corrective actions, licensee self-assessments, and specific aspects of fire protection defense in dept. To the extent practicable, the staff will schedule the triennial inspections so that plants that have not performed or do not plan to perform self-assessments are inspected before those that have done so.

3. Issue the FPFI pilot procedure as a permanent IP. The staff will format the procedure to emphasize its existing modular structure so that individual modules (e.g., fire barriers, fire brigade, arid safe-shutdown capability) could be applied independent of the entire procedure. The FPFI procedure {a) would be used. by the staff to support the triennial inspections as specified in the triennial inspection procedure, (b) would be used by the staff when plant performance falls below a threshold to be established by the new inspection and oversight program or in response to a specific event or problem at a plant, and (c) could be used by the licensees as guidance for self-assessments.

4. pelete IP 64100, IP 64150, and IP 64704 after the new reactor inspection and oversight program is implemented. (The procedures recommended above would supersede these existing procedures. Therefore, IPs 64100. 64150, and 64704 would no longer be needed.)

Staff requests action within 10 days. Action will not be taken until the SRM is received. We consider this action to be within the delegated authority of the EDO.

COORDINATION The Office of the General Counsel has review this Commission paper and has no legal objections.

The Commissioners The Office of the Chief Financial Officer has reviewed this Commission paper for resource implications and has no objections.

William D. Travers Executive Director for Operations Attachments: 1. Fire Risk Fact Sheet

2. FPFI Pilot Program Final Report SECY NOTE: In the absence of instructions to the contrary, SECY will notify the staff on Monda , June 7, 1999 that the Commission, by negative consent, assents to the action proposed in this paper.

DISTRIBUTION:

Commissioners OGC OCAA OIG OPA OCA ACRS CIO CFO EDO REGIONS SECY

Attachment 1 4

FIRE RISK FACT SHEET

~ The average reported fire frequency at operating plants for the period 1965-1994 is 3.3E-1/yr' The average reported fire frequency for the pre-Appendix R implementation period 1965-1985 is 3.8E-1/yr' The average reported fire frequency for the'post-Appendix R implementation period 1985 - 1994 is 2.8E-1/yr'uring the post-Appendix R implementation period (1986-1994) there were two fire events that resulted in a scram and a loss of function of one safety related division or a toss of offsite power. This compares to 10 such events (not including the Browns Ferry, fire) during the pre-Appendix R implementation period (1965-1985).'

There were 41 fire events that resulted in a plant scram with no loss of function of a safety related division in the 20 year pre-Appendix R implementation period and 40 such events in the 8 year post-Appendix R period.'

~ Thirteen large losses from fire events at nuclear power plants during the'eriod from 1966 - 1995 resulted in a total reported monetary loss of approximately $ 800 million, with an average monetary loss per event of approximately $ 62 million.'

On June 21, 1991, the NRC issued GL 88-20, Supplement 4, requesting licensees to perform an Individual Plant Examination of External Events (IPEEE) to (1) develop an appreciation of severe accident behavior, {2) understand the most likely severe accident sequences, (3) gain a guatitative understanding ot the overall likelihood ot core damage and radioactive release, and {4) if necessary, to reduce the overall likelihood of core

'damage and radioactive release by modifying hardware and procedures that would help prevent or mitigate severe accidents.s Special Study Fire Events - Feedback of U.S. Operating Experience, June 1997, James R. Houghton, Office for Analysis and Evaluation of Operational Data, USNRC A 30-Year review of Large Losses in the Gas and Electric Utility Industry - 1966- 1995, James B. Biggins, J&H Marsh & McLennan, 1997 NUREG 1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, USNRC, June 1991

'Page 1 of 2 f

~ Based on the IPEEE results, fire events are important contributors to the'reported core damage frequency (CDF) for a majority of plants. The reported CDF contribution from fire events can in some cases, approach (or even exceed) that from internal events.4

~ The reported IPEEE fire CDFs range on the order of E-9/yr to EA/yr, with the majority of plants reporting a fire CDF in the range from 1E-6/yr to 1EA/yr.'ore than half of the plants proposed or implemented procedural and/or hardware improvements in the fire area in response to their IPEEE.

~ Although most licensees have reported numerical fire CDF estimates, it is important to note that the accuracy of such estimates has not been validated under the IPEEE submittal review. Because simplifying assumptions and approximate procedures may have been used in the analyses, the quantified CDF estimates reported in the licensees'PEEE submittals should ~onl serve as a general indicator of plant risk. With that in mind the following preliminary information is provided.

a. Fire CDFs for approximately 40 units were greater than or equal to 1E-5/yr.

b. Fire CDFs for approximately 11 units were greater than or equal to 1EA/yr.

C. Of those 29 units whose fire CDF was between 1E-5/yr and 1E-4/yr, approximately 17 had a reported fire CDF greater than or equal to the reported internal events CDF.

Of those 11 units whose fire CDF was greater than or equal to 1EA/yr, 9 units had a fire CDF greater than or equal to the reported internal events CDF. For the remaining 2 units the fire CDF was comparable to the internal events CDF.

January 20, 1998, memorandum to the Commissioners from L Joseph Callan, Executive Director for

~

Operations, Preliminary IPEEE Insights Report This range includes all plants except Quad Cities. The licensee for Quad Cities will submit a revised and updated IPEEE fire analysis during May 1999.

Page2of 2

ATTACHMENT2

'ire Protection Functional Inspection Pilot Program Final .Report Plant Systems Branch and Probabilistic Safety Assessment Branch Division of Systems Safety and Analysis Office of Nuclear Reactor Regulation April 5999 L. E. Whitney, P. M. Madden, E. A. Connell, J. S. Hyslop, K. S. West

TABLE OF CONTENTS

1. BACKGROUND . 1 Bases for and Status of the FPFI Pilot Program . '1 FPFI Pilot Program Objectives .3 Scope of FPFI Pilot Program .. .3

2. OVERVIEW OF REACTOR FIRE RISK

3. EXISTING FIRE PROTECTION INSPECTION PROCEDURES

4.

SUMMARY

OF FPFI INSPECTION FINDINGS 8 Pilot FPFls..... 8 River Bend Station =

9 Susquehanna 9 St. Lucie '.. 9 Prairie Island .. .10 Clinton and Quad Cities .. ...

~ 10 Clinton . .10 Quad Cities .11

, 5. FPFI WORKSHOP .12

6. FPFI PILOT PROGRAM INSIGHTS AND LESSONS LEARNED .. .14

7. CONCLUSIONS AND BASES FOR STAFF PLANS .15

8. PLANNED STAFF ACTIONS 17

9. FIRE PROTECTION RISK SIGNIFICANCE SCREENING METHODOLOGY ~ ~ ~ ~ ~ ~ ~ o ~ ~ 18

10. SAMPLE APPLICATIONS OF FPRSSM .. 18 Quad Cities . 19 St. Lucie .............. 19

11. PROGRAM ACCOMPLISHMENTS 20 Appendix A -'FPFI Assessment Tree Appendix B - Fire Risk Fact Sheet Appendix C - Summary of Pilot FPFI Enforcement Actions Appendix D - Summary of Staff Interactions with the Nuclear Energy Institute

Fire Protection Functional Inspection Pilot Program Final Report

1. BACKGROUND Bases for and Status of the FPFI Pilot Pro ram tn a memorandum of August 25, 1992, the staff of the U.S. Nuclear Regulatory Commission

{NRC) submitted to the Commission its action plan for resolving the Thermo-Lag fire barriers issues. The staff stated that it would develop and implement a program to inspect the Thermo-Lag corrective actions at each plant. At that time, the staff believed that the licensees would simply replace or upgrade their existing Thermo-Lag fire barriers. However, since that time, the licensees have proposed a much broader range of corrective action options. For .

example, many licensees have initiated fire barrier reduction programs. The objective of these programs, which are based largely on reassessments and subsequent revisions of the plant post-fire safe shutdown analysis, is to eliminate as much as possible the need for fire. barriers.

Typical outcomes of barrier reduction programs include redefined fire area boundaries, new or relocated safe shutdown components, and new operator actions and procedures. Many licensees were also performing engineering evaluations to justify either eliminating certain Thermo-Lag barriers or keeping them as they are (i.e., without upgrades). In some cases, the licensees used such evaluations to justify exemptions from the NRC fire protection regulations.

ln the memorandum of August 25, 1992, the staff also informed the Commission that it would reassess the NRC reactor fire protection program to (1) determine ifthe program had appropriately addressed the safety issues, {2) determine if licensees are maintaining compliance with the NRC fire protection requirements, (3) identify the strengths and weaknesses of the program, and (4) make recommendations for improvement. The staff issued its "Report on the Reassessment of the NRC Fire Protection Program" on February 27, 1993. That report recommended, in part, that the staff (1) develop'a coordinated approach for the fire protection and systems inspections and {2) reevaluate the scope of the fire protection inspection program. In SECY-93-143, "NRC Staff Actions To Address the Recommendations in the Report on the Reas'sessment of the NRC Fire Protection Program" dated May 21, 1993, the staff informed the Commission that it would implement these reassessment recommendations as part of the Fire Protection Task Action Plan. To do so, the

'staff considered fire events, licensee reports of deficiencies in the fire protection program, previous NRC inspection findings, the scope and adequacy of the existing NRC fire protection inspection program, and the need to inspect other plant fire protection features in response to ongoing NRC programs {e.g., self-induced station blackout, fire barrier penetration seals, turbine building assessments, and individual plant examinations of external events (IPEEEs)).

On the basis of the wide range of Thermo-Lag corrective actions proposed by the licensees, the staff concluded that an inspection of broader scope than that proposed in the Thermo-Lag Action Plan was needed. In addition, in view of the preliminary results of its work under the reassessment recommendation, the staff concluded that additional fire protection inspection effort appeared to be warranted. In SECY-95434, "Status of Recommendations Resulting from the Reassessment of the NRC Fire Protection Program," dated Februaiy 13, 1995, the staff informed the Commission that it was considering initiating a fire protection functional inspection (FPFI) program, which would cover all aspects of nuclear power. plant fire safety

(including Thermo-Lag fire barriers) and provide for more efficient, comprehensive and effective inspections. Revision and/or cancellation of some of the existing fire protection inspection procedures will be considered as part of the FPFI program.

In a memoran'dum to the Commission dated September 20, 1995, the staff documented its conclusion that an inspection of broader scope than that originally specified in the Thermo-Lag Action Plan was needed. The staff also informed the Commission that instead of the stand-alone Thermo-Lag fire barrier inspection program that it had proposed, it would develop and implement the FPFI program it had outlined in SECY-95-034. On February 8, 1996, the staff briefed the Chairman on its plans for the future direction of the NRC reactor fire protection program including the FPFI program. Later, in an April 3, 1996 memorandum to the Commission, the staff documented the framework for future direction of the NRC fire protection program with emphasis on the FPFI program, a plan for developing and implementing this program, and a plan for centralized management, by the Office of Nuclear Reactor Regulation (NRR), of the FPFI program and all other reactor fire protection work.

InPECY-96-267, "Fire Protection Functional Inspection Program," dated December 24, 1996, the staff informed the Commission of its plans for implementing the FPFI pilot program. The proposed FPFI pilot program consisted of developing the FPFI inspection procedure, conducting four pilot inspections and a public workshop, and preparing a final report. In a staff requirements memorandum (SRM) of February 7, 1997, the Commission informed the staff that it did not object to the staff's plans to implement the FPFI pilot program, and indicated its

. interest in strategies that would shorten the time for the benefits of the program to become available to all licensees. The Commission requested that the staff send a report to the Commission at the end of the pilot program. By memorandum dated June 23, 1997, the staff provided its draft FPFI procedure to the Commission.

In SECY-98-'187, "Interim Status Report - Fire Protection Functional Inspection Program,"

dated August 3, 1998, the staff described the inspection results to that date, its plans to complete the pilot program, and the adjustments it had made to the pilot program since it issued SECY-96-267. Later, in SECY-99440, "Second Interim Status Report - Fire Protection Functional Inspection Program," dated February 5, 1999, the staff explained the basis for extending the FPFI pilot program schedule and reported that its final report on the FPFI program would (1) provide an analysis of the'FPFI findings, regional FPFI inspection followup activities, and enforcement actions arising from the pilot FPFls; provide information on the

'. use of risk insights for fire protection inspections; (3) discuss and(2)evaluate the types of NRC fire protection inspections that it has con'ducted since the fire protection regulation was issued in 1981; (4) address the strategies in which the Commission expressed interest in the SRM of February 7, 1997; and (5) recommend the appropriate types and frequencies of reactor fire protection inspection (e.g. NRC-led and licensee self-assessments).

On February 9, 1999, the staff briefed the Commission on the FPFI program. In an SRM of April 1, 1999, the Commission directed the staff to provide its recommendations on the the appropriate types and frequencies of reactor fire protection inspections, and report on whether fire protection functional inspections are needed, and whether fire protection inspections should be part of the new reactor oversight process.

FPFl Pilot Pro ram Ob ectives The objectives of the FPFI pilot program, as stated in the staff documents discussed under Section 1, "Background," are summarized below. Accomplishment of these program objectives is discussed in Section 11 of this report.

s To develop a strong, broad-based, coherent and coordinated NRC Fire Protection Program commensurate with the safety significance of the subject.

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2.. To inspect licensee Thermo-Lag fire barrier corrective actions and compliance strategies.

3. To inspect non-Thermo-Lag fire protection features in response to ongoing NRC programs (e.g., self-induced station blackout, fire barrier penetration seals, turbine building assessments, and lPEEEs).

4. To provide immediate safety benefit through renewed industry attention to nuclear power plant fire safety.

5. To develop criteria for licensee fire protection self-assessments.

6. To ensure compliance with NRC post-fire safe shutdown regulations and commitments.

7. To focus resources on the fire protection issues of most importance.

8. To evaluate the scope and adequacy of the existing NRC reactor fire protection programs, and develop recommendations for program improvements, if warranted.,

9. To review licensee fire protection and post-fire safe shutdown configuration management.

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To provide clear guidance to the staff and industry regarding oversight of reactor fire protection programs.

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11. To address smoke propagation and manual fire fighting operations, and their impact on equipment operability and operator actions.

12. To address balance of plant fire risks.

13. To improve consistency of internal NRC oversight of licensee fire protection program.

14. To address fire safety considerations not expressly addressed by the fire protection regulation (e.g., event based fires, fire-induced plant transients, seismicJfire interactions, and fire-induced release of radioactive materials).

Sco e of FPFI Pilot Pro ram The FPFI pilot program consisted of developing the FPFl inspection procedure, conducting four pilot FPFls and two other FPFl-type team inspections, holding a public workshop, and preparing this final report.

The pilot FPFls were announced, risk-informed inspections that covered all aspects of reactor fire safety. A FPFI assessment diagram is included as Appendix A. The prin'cipai focus of the inspections was on the plant fire protection and post-fire safe shutdown design and licensing bases an'd those fire protection program elements that are covered by existing NRC regulations and guidelines. These included, for example, safe shutdown performance objectives, safe shutdown systems and equipment, fire protection systems and barriers, emergency lighting, reactor coolant pump oil collection systems, quality control and quality assurance, configuration control including change control process, administrative controls and procedures, and training.

This aspect of the FPFI program satisfied the program objective of ensuring continued licensee compliance with NRC fire protection regulations and commitments. In addition, the pilot inspections included reviews of fire safety considerations that are not expressly addressed by the fire protection regulation, but by other regulatory programs. This included, principally, Generic Letter 88-20, Supplement 4, "Individual Plant Examinations of External Events (IPEEE) for Severe Accident Vulnerabilities, 10 CFR 50.54(f)," dated June 28, 1991. SQch inspection areas included, for example, event initiated fires, fire induced reactor transients, and potential seismic fire interactions.

The Office of Nuclear Reactor Regulation (NRR) led pilot FPFls at River Bend, Susquehanna, and St. Lucie using inspectors from NRR, the regional offices, and Brookhaven National Laboratory. NRR coordinated plant selection and inspection schedules with the regional offices. The staff performed the inspections in accordance with the approach described in SECY-96-267 (2 weeks of preparation, 1 week inspecting on site, 1 week reviewing in office, and a final week inspecting on site) using the FPFI procedure that it had sent to the Commission in the memorandum of June 23, 1997. This procedure is much broader in scope than the existing fire protection core inspection procedure (IP 64704, "Fire Protection Program" ). For example, although the objective of IP 64704 is to evaluate the overall adequacy of the licensee's fire protection program, it does not address post-fire safe-shutdown capability, nor does it thoroughly evaluate fire protection program management and configuration control.

The FPFI procedure also differs from the core inspection procedure in that it provides guidance to the inspectors for using risk insights to help focus on areas most important to safety. NRR risk analysts helped the FPFI team obtain fire risk insights for the plant-specific inspection plans.

For these three pilot inspections, NRR prepared the FPFI reports and sent them to the licensees after the appropriate regional office reviewed the report. Like other NRR-led team inspections, and as described in SECY-96-267, the regional offices completed any inspection follow-up and enforcement actions resulting from the FPFI. After NRR issued the FPFI report to the licensee, it made recommendations for inspection follow-up and enforcement to the regional office and supported regional follow-up activities as requested by the region.

Prairie Island was the fourth and final pilot FPFI. In SECY-96-267, the staff stated that licensee self-asseSsments could be an important element of the permanent FPFI program and that it would consider. the role of self-assessments after it completed the pilot FPFI program. In the SRM of February 7, 1997, the Commission stated that it was interested in the use of licensee self-assessments as a strategy to relieve some of the staff inspection burden to the extent that the NRC can be assured that the self-assessments are of good quality and accurately reflect the strengths and weaknesses of the program. The Commission noted that staff review of the self-assessments would be warranted to gain this assurance. After the staff announced the FPFI pilot program, Northern States Power Company conducted a self-assessment of the

Prairie Island fire protection and post-fire safe-shutdown programs in anticipation of receiving a pilot FPFI. This gave the staff an opportunity to test an inspection strategy involving licensee self-assessments as part of the FPFI pilot program. This pilot inspection differed from the three previous pilot inspections in two significant ways, First, it was a reduced-scope inspection of a licensee self-assessment instead of a full-scope FPFI. Second, it was led by the region rather than by NRR. NRR provided staff and contractor support to the region.

ln contrast to a full FPFI, the self-assessment inspection of Prairie.Island was a one-week inspection. The NRC inspection team evaluated the licensee's self-assessment effort to determine whether or not the scope and depth of the effort were equivalent to an FPFI, or if the licensee had an acceptable basis, for reducing the scope or depth. The team reviewed the licensee's organization, the technical qualifications of the licensee's assessment team, the completeness of the assessment, the corrective actions proposed by the licensee for the more significant assessment findings, and the licensee's handling of any operability concerns. The staff considered this exercise in formulating its recommendations for future reactor fire protection inspections.

During the FPFI pilot program period, the staff also conducted major team inspections at Clinton and Quad Cities. Like the FPFI pilot inspections, these inspection experiences provided insights into possible weaknesses with the core fire protection inspection program, the potential benefits of more comprehensive fire protection inspections (like FPFIs), and the use of licensee self-assessments. Therefore, the staff also considered the results of these two inspections when it assessed the insights and lessons learned from the FPFI pilot program and developed its recommendations for the types and frequencies of future reactor fire protection inspections.

The findings from the pilot FPFls and from the team inspections at Clinton and Quad Cities are summarized in Section 4 of this report.

On November 10, 1998, the staff held a onMay workshop on reactor fire protection inspections. The workshop is discussed in Section 5 of this report.

2. OVERVIEW OF REACTOR FIRE RISK

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Risk assessments have shown that fires in nuclear power plants can be risk significant. To ensure that its decisions and recommendations regarding the regulatory program for reactor fire protection are commensurate with its importance and risk significance, the staff considers the underlying purpose of the regulatory requirements and, to the extent practicable, available fire risk data and insights.

Simply stated, the underlying purpose of the NRC fire protection regulation is to provide reasonable assurance that one means of achieving and maintaining safe-shutdown conditions will remain available during and after a fire. This is accomplished by applying the concept of fire protection defense in depth to reduce the likelihood of fires and to limit the extent of fire damage to the structures, systems, and components that would be used to achieve and maintain safe shutdown in the event of a fire. The objectives of defense in depth are to prevent fires from starting; to rapidly detect, control, and extinguish fires that do start; and to design and protect structures, systems and components so that a fire that is not promptly extinguished will not adversely affect safe shutdown.

Fire is not treated as a design-basis accident, nor are fires postulated to occur simultaneously with non-fire-related failures in safety systems, plant accidents, or severe natural phenomena.

The regulation requires that only one train of equipment necessary to achieve hot shutdown be maintaine'd free of fire damage. Unlike systems set up to mitigate design-basis accidents, the fire protection regulation does not require redundant or diverse post-fire safe-shutdown methods. Nor does the regulation require that fire protection systems and features or the structures, systems, and components provided for achieving post-fire safe shutdown be safety related, protected against a single failure, or covered by technical specifications. Finally, the regulation does not require that the equipment provided to achieve cold shutdown or to mitigate

'he consequences of design-basis accidents be maintained free of fire damage.

In 1989, Sandia National Laboratories issued the "Fire Risk Scoping Study." This study, which included a review of the fire probabilistic risk assessments (PRAs) for four plants, concluded that the most risk significant plant areas typically are main control rooms, cable spreading

. rooms, and switchgear rooms. According to this study, although plant modifications made in

, response to Appendix R reduced, by a factor of 3 to 10, the core damage frequencies (CDFs) at the plants studied, fire can be an important contributor to CDF even after regulatory criteria have been satisfied. The study suggests that without the existing regulatory requirements, fire risk could be higher than it is today. The study also suggests that improper implementation of regulatory requirements and degradation of fire protection defense'in depth could be risk

'he significant. The study concluded, for example, that weaknesses in either manual fire fighting effectiveness or control systems interactions could raise the estimated fire-induced CDF by an order of magnitude.

Under the IPEEE program, the licensees systematically assessed the fire risk for each operating reactor. The results of the IPEEEs confirm the results of the "Fire Risk Scoping Study." Although most licensees have reported numerical fire CDF estimates, the staff has not validated the accuracy of such estimates.. Because the licensees may have used simplifying assumptions and approximate procedures in the analyses, the quantified CDF estimates reported in the IPEEE submittals only serve as general indicators of plant fire risk.

Nevertheless, the IPEEE fire analyses provide important insights regarding reactor fire risk. For example, the IPEEE results show that fire events are important contributors to the reported CDF for a majority of plants, ranging on the order of1 E-9/yr to 1EA/yr, with the majority of plants reporting a fire CDF in the range of 1E-6/yr to 1E-4/yr. The reported CDF contribution from fire events can in some cases approach (or even exceed) that from internal events. More than half of the plants proposed or implemented procedural and/or hardware improvements in the fire area in response to their IPEEE. (In most cases any risk reduction achieved by the improvements was not reported separately, but was included in the total CDF.) Overall, the IPEEEs have confirmed that main control rooms, cable spreading rooms, and switchgear rooms.

are usually the most risk-sighificant plant areas, Although it is generally understood that fire events can be serious and risk significant, fire science is a relatively new field. NRC fire research efforts and fire risk assessments have yielded both useful tools and important results. However, a number of important questions remain regarding the assessment and assurance of nuclear fire safety. For example, there are still significant uncertainties in the ability to mechanistically predict the behavior of fires under the broad variety of conditions that are relevant to nuclear power plant safety..

Appendix B, "Fire Risk Fact Sheet," gives additional information about reactor fire risk.

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3. EXISTING FIRE PROTECTION INSPECTION PROCEDURES As discussed under Section 2, the underlying purpose of the NRC fire protection regulation is to provide reasonable assurance that one means of achieving and maintaining safe shutdown conditions will remain available during and after a fire. This is accomplished by applying the concept of fire protection defense-in-depth to reduce the likelihood of fires and to limit the extent of fire damage to the structures, systems, and components that would be used to achieve and maintain safe shutdown in the event of a fire. Since the fire protection regulation was issued in 1981, the staff published three reactor fire protection inspection procedures (IPs).

These are IP 64100, "Postfire Safe Shutdown, Emergency Lighting and Oil Collection Capability at Operating and Near-Term Operating Reactor Facitities"; IP 64150, "Triennial Post-Fire Safe Shutdown Capability Reverification"; and IP 64704, "Fire Protection/Prevention Program."

During the 1980s, the staff inspected each plant one time using IP 64100, a 1-week, region-led team inspection. The overall objective of IP 64100 was to verify, through a sampling audit at a specific reactor facility, that there was reasonable assurance that the facility could achieve and maintain safe shutdown in the event of a fire, IP 64100 inspections generally consisted of a fire brigade drill, walkdowns of post-fire safe shutdown equipment locations and fire area and/or fire zone configurations, reviews of post-fire safe shutdown procedures, a post-fire safe shutdown procedure simulation, and audits of fire-induced circuit failure analyses. Typically, the inspectors verified the existence of fire detection systems, fire, suppression systems, and fire barriers installed to protect post-fire safe shutdown equipment. However, the inspectors did not have sufficient time to inspect details of.the safe shutdown analysis or the design basis of fire detection systems, fire suppression systems, and fire barriers.

After the original post-fire safe shutdown inspections (IP 64100), the regional initiative, one week, triennial team inspections under IP 64150, were intended to venfy continued compliance with regulatory requirements regarding post-fire safe shutdown. IP 64150 was rarely conducted due to the lack of resources such as allocated inspection hours and the availability of experienced and skilled post-fire safe shutdown mechanical systems and electrical engineering specialists. Since the one-time IP 64100 inspections, the staff has inspected fewer than 10 plants in accordance with IP 64150.

Throughout the 1980s and 1990s the regions have conducted IP 64704, the routine (core) reactor fire protection inspectiori procedure at each plant about once every 3 years. The objective of this IP, which is typically conducted by a regional inspector over about a 1-week period, is to evaluate the overall adequacy of the licensee's fire protection program. IP 64704 focuses on such "classical" fire protection features as administrative controls of transient combustibles and ignition sources, surveillance and testing of detection and suppression systems, fire barrier integrity, quality assurance audits, and general employee and fire brigade fire response training and capabilities. IP 64704 does not address post-fire safe-shutdown capability,'or does it thoroughly evaluate the overall fire protection program configuration control. It should be noted that licensee QA audits or self-assessments that are based on IP 64704 would similarly not address post-fire safe shutdown capabilities.

As discussed in SECY-93-143, "NRC Staff Actions To Address the Recommendations in the Report on the Reassessment of the NRC Fire Protection Program," May 21, 1993, on the basis of its self-assessment, the staff concluded that the existing NRC fire protection requirements

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would, if properly implemented, effectively address fire safety concerns at commercial nuclear power plants. The staff acknowledged, however, that it had not been inspecting post-fire safe shutdown capabilities. On the basis of its self-assessment, and in light of the Thermo-Lag fire barrier is'sue, the staff concluded that the NRC staff and the licensees had reduced their emphasis on fire protection after the licensees achieved compliance with fire protection requirements and the staff completed its one-time, post-fire, safe shutdown inspections. The staff initiated the FPFI pilot program, in large part, to assess the impact of this reduced licensee

'nd regulatory emphasis.

4.

SUMMARY

OF FPFI INSPECTION FINDINGS The findings from. the pilot FPFls and the team inspections at Clinton and Quad 'Cities are summarized below. These inspections were intended to uncover problems and vulnerabilities in the licensee's fire protection programs. Therefore, this report section highlights fire protection and post-fire safe shutdown program deficiencies and degradations rather than strengths.

Pilot FPFI enforcement'actions are listed in Appendix C.

Pilot FPFls At all FPFI pilot plants, fire brigade equipment, fire fighting strategies, and performance were deficient to some degree. Problem areas included personnel safety equipment, equipment staging locations, response time, limited offsite'fire department resources, communications, medical examination frequency, lack of flammable liquid suppression capability, drill realism, fire fighting water drainage control, smoke removal, radiological control, and inadvertent or planned breaching of redundant train fire barriers as part of the fire fighting strategy.

Other pilot FPFI inspection findings were in the areas of: control of combustibles; fire detector and sprinkler system design (failure to meet minimum code criteria); gaseous suppression system acceptance testing and hazard suitability; hose and standpipe coverage and sunreillance criteria; deficient fire area. boundaries due to maintenance, installation defect or design; inadequate safe shutdown passive barrier separation for redundant components; ineffective compensatory measures for removed fire barriers; inability of selected safe shutdown systems to meet performance goals given credible fire-induced actuations; inadequate electrical safe shutdown analyses (e.g., inappropriate assumptions that multiple spurious operations would not occur from a fire in a single fire area, common power supply concerns, and potential motor operated valve (MOV) mechanical damage due to fire-induced

,spurious operations which may bypass valve limit or torque switches); inadequate attention to fire protection program management; incomplete alternative safe shutdown procedures; inadequate emergency lighting to accomplish required operator actions; inadequate or potentially fire damaged communications capabilities; and nonwonservative or invalid IPEEE assumptions.

Summaries of the pilot FPFI findings by plant are presented below.

River Bend Station The results of the, River Bend Fpi=l are documented in NRC Inspection Report Number 50-458/97-201'ated March 20, 1998. The inspection team reported the following findings:

(1) there was a weakness in how transient combustibles are controlled; (2) smoke detection and fire suppression system designs did not meet industry standards; (3) there was a weakness in the analysis and testing of fire doors; (4) engineering evaluations of certain fire barrier designs did not demonstrate that the barriers protected adequately. against the fire. hazards; (5) fire brigade performance was weak; (6) compensatory measures for the lack of certain fire barriers did not provide an equivalent level of safety; and (7) certain IPEEE assumptions were weak. The inspection team also found that the licensee's post-fire safe- shutdown circuit failure analysis methodology did not consider multiple circuit faults and, therefore, did not identify certain conditions that could prevent the operation or cause the maloperation of post-fire safe-shutdown capability (e.g., a potential fire-induced reactor transient may not have been properly analyzed and bounded). As part of its Thermo-Lag corrective action program, the licensee re-analyzed its post fire safe shutdown methodology. The objective of the re-analysis was to reduce reliance on Thermo-Lag fire barriers and to upgrade required Thermo-Lag barriers. The inspection team did not identify any problems with the licensee's Thermo-Lag corrective action program.

'usquehanna The results of the Susquehanna FPFI are documented in NRC Inspection Report Nos. 50-387/97-201 and 50-388/97-201 dated May 13, 1998. The inspection team reported the following findings: (1) transient combustibles were not,controlled in accordance with plant procedures; (2) the fire brigade drill revealed response and firefighting technique problems; (3) fire detection and suppression system designs did not meet fire protection industry codes and standards; (4) the post-fire safe shutdown method for certain fire areas used the automatic depressurization and core spray systems and could allow core uncovery; (5) off-normal post-fire safe-shutdown procedures did rot fully identify all required manual actions or did not identify the preferred instrumentation to be used to monitor reactor performance; and (6) emergency lighting was not provided for certain safe-shutdown operations. The inspection team noted that licensee personnel exhibited good. knowledge of the Susquehanna fire protection features and post-fire safe-shutdown capability, that the scope and depth of operator training was good, that the licensee had been pro-active in addressing Kaowool fire barrier concern, and that modifications had been implemented to prevent fire-induced spurious actuations of motor-valves (MOVs). During the inspection the licensee was in the process of confirming 'perated the design attributes of the installed Thermo-Lag fire barriers and evaluating required barrier upgrades. The inspection team did not identify any problems with the licensee's Thermo-Lag corrective action program.

St. Lucie

'he results of the St. Lucie FPFI are documented in NRC Inspection Report Nos. 50-355/98-201 and 50-389/98-201 dated July 9, 1998. The inspection team reported the following findings: (1) fire detection and suppression system designs did not meet fire protection industry codes and standards; (2) transient combustibles were not controlled in accordance with plant procedures; (3) the fire brigade drill performance revealed response and firefighting technique problems; (4) the safe-shutdown analysis did not consider the fire-induced affects of multiple 9

high-impedance electrical faults associated with the power distribution system; (5) weaknesses were associated with the fuse breaker coordination control program; (6) there were no fire isolation measures to protect against fire-induced spurious operation of high/low reactor pressure boundary valves; (7) there was no fire barrier to separate post-fire safe-shutdown charging function; (8) there was no analysis of fire-induced affects on instrument sensing lines; (9) there was a potential for fire-induced circuit failures leading to spurious operation of required post-fire safe shutdown MOVs; and (10) there was no emergency lighting for certain post-fire safe-shutdown operations. The team also found that there was a general lack of fire protection and post-fire safe-shutdown program ownership by the engineering department, and that the licensee's response to negative quality assurance findings was slow. With respect to the licensee's Thermo-Lag fire barrier upgrade program, the inspection team found that certain wall upgrades and designs were not sufficient to provide the fire resistance needed to contain the fire hazards in the areas of concern and that adequate fire resistive protection was not provided for thermal shorts that penetrate Thermo-Lag raceway fire barriers.

Prairie Island The results of the Prairie Island FPFI are documented in NRC Inspection Report Nos. 50-282/98-016 and 50-306/98-016 dated October 9, 1998. At Prairie Island, the NRC conducted a reduced-scope, one week inspection to review and validate a licensee fire protection program self-assessment. The inspectors concluded that the licensee's self-assessment process was acceptable. However, inspection findings included: omission of eight residual heat removal (RHR) containment sump suction valves from the analysis; missing one-hour rated fire barriers on electrical conduits for an auxiliary feedwater (AFW) pump suction valve; indeterminate fire-resistive performance of the Kaowool fire barrier system in the plant; and weakness in the safe

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shutdown timeline analysis for a fire-induced transient. Of special significance were instances in which the licensee had removed 1-hour fire-rated barriers that it had committed to maintain as part of approved Appendix R exemptions. In addition, through its self-assessment, the licensee found: 32 safe shutdown-related motor operated valves (MOVs) that were susceptible to fire damage due to fire-induced hot shorts; several inadequate Appendix R fire barriers and unsealed fire barrier penetrations; pressurizer level indication channels that did not meet .

separation requirements; and a 1-hour fire-rated barrier missing from a safety injection (Sl) pump suction valve electrical conduit.

Clinton and Quad Cities The staff did not conduct FPFIs at either Clinton or Quad Cities. However, as discussed in the following sections, recent experiences with these plants provided insights into possible weaknesses with the core fire protection inspection program, the potential benefits of more comprehensive fire protection inspections (like FPFls), and the use of licensee self-assessments.

Clinton The results of the Clinton fire protection related inspections are documented in a Special Evaluation Team (SET) report dated January 2, 1998, and Inspection Report Nos. 5(F461/98-026 dated December 13, 1998. The results of a related follow-up fire protection inspection will be documented in Inspection Report 50-461/99-003 to be issued. The staff had scheduled a pilot FPFI at Clinton Power Station. In preparation for the FPFI, the licensee performed an augmented fire protection program quality assurance audit and found that a program breakdown existed. The licensee issued 16 condition reports, 11 of which were attributed to inadequate post-fire safe-shutdown analyses. Before the staff could perform the pilot FPFI, it was canceled due to the licensee's commitment to perform an Independent Safety Assessment (ISA) and the NRC's oversight of this effort with a Special Evaluation Team (SET). Because of the significance of the licensee's fire protection audit findings, the SET performed an in-depth, vertical-slice inspection of the Clinton fire protection program. The SET noted that the licensee could not demonstrate the ability of the post-fire safe-shutdown analysis, equipment, and procedures to ensure that the plant could achieve and maintain safe-shutdown following a fire.

In its report, the SET reported that: the licensee had failed to ensure that 54 MOVs would remain free of fire damage due to fire-induced hot shorts in the valve control circuitry; the licensee was unable to provide acceptable industry fire barrier test reports to support the upgrade of an installed Thermo-Lag fire barrier assembly; the licensee did not have adequate test documentation to support the qualification of fire rated cables as equivalent to'that of a fire rated barrier; and that the sprinkler systems in several risk significant fire areas may be incapable of suppressing a fire.

This experience demonstrated how one of the proposed benefits of the FPFI program, to gain renewed industry attention to nuclear power plant fire safety, is to be achieved. That is, implementation of the FPFI program led the licensee to assess its fire protection program, revealing significant programmatic and fire safety issues. The staff also notes that routine fire protection core inspections had not and would not have uncovered many of the issues that the licensee identified in preparation for the FPFI, but an FPFI-type inspection would have done so.

This experience also produced insights into the possible benefits and uses of licensee self-assessments as a reactor fire protection inspection strategy.

Quad Cities The results of the Quad Cities fire protection inspection are documented in Inspection. Report Number 50-254/98-011 dated July 2, 1998. In September 1997, the licensee found problems with the Quad Cities post-fire safe-shutdown procedures and declared all safe-shutdown paths inoperable. In December 1997, after significant effort to correct these and other fire protection the licensee was unable to demonstrate to the staff that the Quad Cities safe- 'roblems, shutdown analysis and procedures were adequate to assure that a fire in any plant area would not prevent the performance of necessary post-fire safe-shutdown functions. Ultimately, the licensee shut down both units to address these problems. Later, after a fire protection-related restart team inspection, th'e units restarted.

At Quad Cities the licensees alternative post-fire safe shutdown capability had a number of significant weaknesses. These included: loss of control room indications; a forced dual unit station blackout condition; the potential for secondary fires which could interfere with timely safe shutdown', inadequate determination of the worst-case spurious signal or operation, inappropriate assumption that only one spurious operation would occur as a result of a fire in any area, and inappropriately crediting automatic actions for main steam isolation valve closure.

In addition, at Quad Cities a turbine building common area did not appear to have adequate detection and suppression equipment to assure alternative post-fire safe shutdown, the safe

'shutdown analysis did not adequately consider the smoke hazards to operators who would implement post-fire alternative safe shutdown measures, timeline and environmental condition considerations within the safe shutdown procedures were marginal, safe shutdown analysis

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review documentation was lacking, communication systems minimally supported the complex dual unit shutdown and fire brigade operations, administrative procedures to assure non-unit specific safe shutdown equipment operability needed enhancement. The licensee did appropriately identify the systems needed to achieve and maintain post-fire safe shutdown stage sufficient quantities of good material condition manual fire suppression 'onditions, equipment, and provide adequate 8-hour emergency lighting for required safe shutdown activities.

This experience is another example in which the core inspection had not and would not have revealed significant fire safety issues, but an FPFI would have. It also provided insights into the possible benefits,and use of licensee self-assessments as a reactor fire protection inspection strategy.

5. FPFI WORKSHOP On November 10, 1998, the staff sponsored a one-day workshop on reactor fire protection inspections. More than 170 people attended, about 150 worked for licensees, architect-engineer and consulting firms, such industry organizations such as NEI and the Boiling Water Reactor Owner's Group (BWROG), the press, and intervenor groups, and the rest were NRC staff from NRR, the Office for Analysis and Evaluation of Operational Data, the Office of Research, the Advisory Committee on Reactor Safeguards, and each of the four NRC regional offices.

The main purpose of the workshop was to discuss with the stakeholders options for the future

'irection of NRC reactor fire protection inspections in light of the lessons learned from the FPFI pilot program. NRC staff and managers made presentations on the overall direction of performance assessment and NRC inspections, the results of the four FPFI pilot inspections, the results of FPFI-like insp'ections at the Quad Cities and Clinton, the types and frequencies of the inspection findings, the risk significance of the findings, and considerations for developing options for the future direction of the NRC reactor fire protection inspection program.

During the workshop, the staff noted that about one quarter of the findings were in the area of post-fire safe shutdown, an area not covered by the NRC core fire protection inspection procedure or licensee audits thatare based on the core procedure. The staff also noted that one of the proposed benefits of the FPFI program, renewed industry attention to nuclear power plant fire safety, had been achieved. Several workshop participants, including NEI, expressed agreement with this assessment. The staff also noted that routine fire protection core inspections had not and would not have found many of the issues that the licensees identified in preparation for FPFls and during self-assessment, or that the staff found during FPFls. Finally, the staff noted that the FPFI pilot program provided insights into the possible benefits and uses of licensee self-assessments as a reactor fire protection inspection strategy.

Staff and industry representatives also discussed the use of risk techniques and insights for fire protection inspections, for example, for planning lines of inspection inquiry and assessing the risk and safety significance of inspection findings. There was general agreement that the tools to measure the risk significance of spec Tiic fire protection inspection findings are not mature and that relying on the results of IPEEE needs to be carefully considered because of the assumptions that go into the analyses and the screenings.

Representatives of two licensees of reactor plants at which pilot FPFI inspections were conducted discussed their experiences. NEI presented the results of an industry-survey quantifying the frequency and focus of. recent licensee self-assessments. NEI presented a preliminary proposal for an industry initiative to strengthen the self-assessment process using ~

FPFI procedures and techniques.

Key messages expressed by stakeholders at the workshop included the following:

1. NRC FPFI activities have heightened industry awareness of fire protection and post-fire safe shutdown issues, the importance of fire protection programs, and the need for licensee self-assessments in this area. It is important to have strong, focused,,in-depth, cost-effective fire protection and safe shutdown programs. Re-examination of commercial nuclear power plant fire protection along the lines of the pilot FPFls is needed.

2. Some commercial nuclear plant fire protection programs do contain weaknesses today, in due to a lack of industry awareness of important issues in this tech'nical area. Some

'art licensee representatives expressed uncertainty as to what constitutes compliance with the present fire protection requirements. This uncertainty arises due to issue complexity, technical ambiguities, multiple interpretations, numerous and sometimes vague documents (e.g., Branch Technical Positions, Generic Letters, Bulletins, Information'Notices, Safety Evaluation Reports, etc.), and changing expectations.

3.~ A consistent tool for assessing the safety and risk significance of fire protection inspection findings is needed. The staff's response to this issue is presented in Section 9 of this

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report. ~

4. The first three major team pilot FPFls were conducted over a too short penod of time given their broad scope and inMepth approach, were too resource intensive, and were costly to support, requiring augmentation by contractors and NE firms (even though licensee preparation efforts should have been mainly comprised of revalidations of existing capabilities). Some expressed the view that the FPFls were not cost effective relative to the safety significance of the inspection findings. One licensee representative stated that narrowly focusing on licensees'ire protection program configuration management under 10 CFR 50.59 would be more efficient than the broad scope FPFls.

5. Fire protection and safe shutdown self-assessments have been common in the nuclear industry since 1997, but not all licensees have conducted them and not all of those self-assessments have been broad and in<epth. There are still questions within industry as to what constitutes compliance with Appendix R requirements.

6. The risk-informed, draft FPFI inspection procedure was accepted as a source of appropriately structured, effective and efficient lines of inspection inquiry, and also a source of insightful and detailed inspection guidance.

7. NEI proposed, as one possible future course of action, a fire protection and post-fire safe shutdown re-examination model consisting of licensee-managed self-assessments.

8. NEI stated that systematic licensee fire protection/safe shutdown self-assessments should be integrated into the new reactor inspection and oversight program.

A summary of staff interactions with the Nuclear Energy Institute since the November 10, 1998 workshop on reactor fire protection inspections is provided in Appendix D.

6. FPFI PILOT PROGRAM INSIGHTS AND LESSONS LEARNED As discussed during the Commission meeting on February 9, 1999, one of the proposed benefits of the FPFI program, to gain renewed industry attention to nuclear power plant fire safety, was achieved. For example, in response to the FPFI pilot program, a number of licensees conducted comprehensive self-assessments of their fire protection programs even though they had not been selected as pilot plants.

Since. the staff completed its original post-fire safe-shutdown inspections (IP 64100), many licensees made major changes to their NRC-approved fire protection and post-fire safe-shutdown programs. For example', to resolve Thermo-Lag issues, licensees redefined plant fire area bouridaries, removed fire barriers, rerouted cables and relocated equipment, changed safe-shutdown methods, and added new operator actions and procedures. Under the current regulatory framework, many of these changes have not been subject to NRC review or inspection.

For each FPFI, a senior NRR risk analyst reviewed available IPEEE results and other sources of risk information. (The FPFI inspection procedure contains guidance for using risk information and insights to focus inspection activities.) The risk insights, which were used as input to the FPFI inspection plans, helped focus the FPFls on areas in which the potential fire risks were greater and helped the inspectors improve their understanding of the inspection findings.

As discussed in SECY-98-1 87, potentially risk significant FPFI findings related to the regulatory requirements and licensee commitments had not been and would not have been revealed using the current core fire protection inspection procedure (IP 64704). Similarly, licensee quality assurance audits of reactor fire protection progams had not uncovered many of the findings related to the regulatory requirements and licensee commitments that were revealed during the pilot FPFls.

Until the FPFI pilot program, the staff did not inspect the design basis of fire detection systems, fire suppression systems, and fire barriers installed to protect safe shutdown equipment. The FPFls revealed a number of findings in these areas, including actions taken by licensees to resolve Thermo-Lag fire barrier issues. Because of the importance of these defense-in-depth features wIth'respect to protecting the safe-shutdown ca'pability in the event of a fire, some level of NRC inspection is warranted.

Although'the FPFI program involved a relatively small sample of plants, the inspection results indicate weaknesses in licensee fire protection and post-fire safe-shutdown programs. The FPFI pilot program results suggest that deficiencies could exist in one or more layers of fire protection defense in depth at any given plant, and that the deficiencies could be risk significant. For example, all of the pilot plants had some fire brigade weaknesses. Other findings were, for example, inadequate safe shutdown analyses, incomplete safe shutdown procedures, and inadequate attention to fire protection program management.

The licensees for several of the FPFI pilot plants had conducted fire protection program self-assessments in advance of the FPFI. One of the pilot FPFIs was a reduced-scope inspection of a licensee self-assessment. The self-assessments were based largely on the FPFI procedure'and lessons learned by the licensees by observing previous pilot FPFls. Overall, the licensee self-assessments were of good quality, were commensurate with an FPFI, and reflected the strengths and weaknesses of the licensees'rograms fairly well. On the basis of its specific experience associated with the Prairie Island self-assessment inspection, the staff concluded that (1) follow-up inspections of licensee fire protection program self-assessments could be a satisfactory means of fire protection program oversight, (2) the FPFI pilot inspection procedure (TI 2515/XXX) is adequate and appropriate to guide the conduct of followup inspections of licensee self-assessments, (3) it is important that the licensee self-assessment audit plan, focus at a minimum on each of the topical areas addressed in the FPFI inspection procedure, (4) the licensee should have finalized all technical analysis and documentation before the NRC inspection team visits the site to gather information needed to prepare for its

. inspection, and (5) the quality and completeness of the licensee's self-assessment program documentation significantly affect the scope and depth of the inspection team's activities.

7. CONCLUSIONS AND BASES FOR STAFF PLANS On the basis of the insights and lessons learned from the FPFI pilot program, the importance of fire protection from the point of view of potential risk, past operational experience (e.g., such issues as Thermo-Lag and circuit analysis), and the regulatory requirements, the staff concludes that it should continue some level of inspection of reactor fire protection programs.

The staff believes that future fire protection inspections should be more risk informed than current core inspections and should include the post-fire safe shutdown capability, which is not covered by the current core inspection program. The staff also concludes that intense fire protection inspections, such as full-scope FPFls, are not warranted as a routine-type inspection, but the FPFI procedure should be available for use on an as-needed basis, such as, when plant performance declines;- or to respond to a specific event or problem at a plant. The staff also concludes that licensee self.-assessments should be considered during future NRC inspections, provided that the scope and depth of the self-assessments 'are at least equivalent to that of the NRC inspections discussed below. In such cases, the NRC inspections would verify the accuracy of the licensees'ssessment and review the licensees'ffectiveness in finding and resolving problems. Finally, the staff concludes that future NRC fire protection inspections should be consistent with the concepts and objectives of the new reactor inspection and oversight program and should be included within that program. Specific staff plans are presented below.

Within the new reactor inspection and oversight program, fire protection defense-in-depth is addressed by the initiating events cornerstone (combustible material and ignition source control) and the mitigation systems cornerstone (fire detection and suppression features, and post-fire safe shutdown capability). Fire protection is not covered by performance. indicators. In SECY-99-007A, "Recommendations for Reactor Oversight Process Improvements (Follow-Up to SECY-99-007)," March 22, 1999, the staff informed the Commission that it had drafted procedures for baseline inspections of fire protection programs for use under the new reactor inspection and oversight program. This baseline inspection procedure is the foundation of the staff's plans regarding the appropriate types and frequencies of reactor fire protection inspections and whether they should be part of the new reactor oversight process.

The baseline procedure calls for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per month of routine resident inspector assessment of fire detection and manual and automatic suppression capabilities, barriers to fire propagation, and fire protection related compensatory measures. The resident inspectors'ssessments of the licensees'ontrol of transient combustibles and ignition sources would be addressed on a more frequent basis in the plant status inspection procedure.

The proposed reactor fire protection baseline inspection procedure also specifies a risk-informed, trienniel, one week, team inspection of each licensee's fire protection program. This triennial inspection will involve a 2-3 day information gathering site visit, and will be conducted onsite by a team comprised of a fire protection engineer, a mechanical engineer, and an electrical engineer. The triennial team inspection is intended to look at all three elements of fire protection defense-in-depth, with major emphasis on post-fire safe shutdown capability and

'configuration management. The inspectors would selectively adopt, as appropriate, inspection techniques developed during the FPFI pilot program. A senior reactor analyst (SRA) would provide input on risk insights for the inspection plan.

The staff also proposes that the FPFI pilot procedure be issued as a permanent IP for use under the new reactor inspection and oversight program. The staff would format the procedure to emphasize its existing modular structure such that individual modules (e.g., fire barriers, fire brigade, safe shutdown capability, etc.) could be applied independent of the entire procedure.

The FPFI procedure (1) would be used by the staff to support the triennial inspections as specified in the triennial inspection proc'edure, (2) would be used by the staff when plant performance falls below a threshold to be established by the new inspection and oversight program or in response to a specific event or problem at a plant, and (3) could be used by the licensees as guidance for self-assessments.

The proposed fire protection baseline procedure, and the staff's plans to issue the FPFI inspection procedure to support the assessment of reactor fire protection programs under, the new reactor inspection and oversight program, will accomplish the following:

~ Prioritization of the FPFI "modules" so that most significant parts of the inspection can be included along with other staff inspections. As part of the new reactor oversight and inspection program the staff will reformat the supporting FPFI procedure to further emphasize its existing modular structure. Then, if needed, individual modules could be independently performed during other types of NRC inspections.

~ The strategy of using licensee self-assessments to relieve some of the staff inspection

'urden to the extent that the NRC can be assured (through inspection) that the self assessment is of good quality and accurately reflects the strengths and weaknesses of the program would be met by baseline procedure direction to the triennial inspection team leader that the inspection plan take into account self-assessment results, taken in combination with NEls expectation that licensee self-assessments will be phased-in during summer 1999.

~ The risk-focused triennial team inspections would be guided by the proven inspection techniques of the FPFI pilot inspection procedure. These triennial inspections will be more narrowly focused than the full-scope pilot FPFI inspections, but their lines of inspection inquiry will be individually tailored during succeeding triennial inspections through the selection of modules from the FPFI procedure. The triennial baseline inspections will,

-1 6-

therefore, validate licensee fire protection and post-fire safe shutdown configuration-management activities.

~ Triennial team inspections would enhance nuclear power plant fire safety by heightening licensee awareness of the importance of fire protection programs and encouraging licensee self-assessments.

The SRM of February 7, 1997, stated that the Commission understood that the staff was considering prioritization of plant reviews so that the most vulnerable plants are reviewed first.

ln that time frame, it was envisioned that from four to eight full-scope FPFls would likely be conducted each year. However, with the development of the new reactor inspection and oversight program a'nd its proposed baseline triennial fire protection inspections, and the staff's plans to conduct 1-week team inspections at each site every 3 years, the concern has lessened that vulnerable plants will be overlooked. Such triennial inspections would be scheduled under priorities set within'the reactor oversight and inspection program.

8. PLANNED STAFF ACTIONS Unless otherwise directed by the Commission, the staff will:

Include risk-informed baseline procedures for routine resident inspector walkdowns and for triennial fire protection team irispections within the new reactor inspection and oversight program to monitor licensee performance in the fire protection area. The staff previously described this approach in. SECY-99-007A.

2. Structure the triennial baseline inspection procedure to emphasize its modular nature so that it could be used, for example, to independently inspect Thermo-Lag corrective actions, licensee self-assessments, and specific aspects of fire protection defense in depth. To the extent practicable, the staff will schedule the triennial inspections so that plants that have not performed or do not plan to perform self-assessments are inspected before those that have done so.

3. Issue the FPFI pilot procedure as a permanent IP. The staff will format the procedure to emphasize its existing modular structure so that individual modules (e.g., fire barriers, .

fire brigade, and safe-shutdown capability) could be applied independent of the entire procedure. The FPFI procedure (a) would be used by the staff to support the triennial inspections as specified in the triennial inspection procedure, (b) would be used by the staff when plant performance falls below a threshold to be established by the new inspection and oversight program or in response to a specific'event or problem at a plant, and (c) could be used by the licensees as guidance for self-assessments.

4. Delete IP 641 00, IP 641 50, and IP 64704 after the new reactor inspection and oversight program is implemented. (The procedures recommended above would supersede these existing procedures. Therefore, IPs 64100. 64150, and 64704 would no longer be needed.)

9. FIRE PROTECTION RISK SIGNIFICANCE SCREENING METHODOLOGY 1

At the time of the FPFI pilot inspections, a tool for systematically assessing the risk significance of fire protection inspection findings was not available. During the FPFI workshop, there consensus that such a tool would be beneficial to both the staff and the industry. was'eneral Subsequently, NRR's Plant Systems Branch and the Probabilistic Safety Assessment Branch, with assistance form the Office of Nuclear Regulatory Research and the senior reactor analysts, developed a proposed method for assessing the potential fire risk significance of fire protection inspection findingsThe Fire Protection Risk Significance Screening Methodology (FPRSSM) approximates core damage frequency (CDF) changes resulting from fire protection program deficiencies. After it is completed, the FPRSSM could be used by inspectors to focus on risk-significant sets of inspection findings, while screening out findings that have minimal or no risk significance. The FPRSSM could also be used to evaluate inspection findings after the inspection. Inspection findings that the FPRSSM finds to be potentially risk significant (i.e.,

those that are not screened out as having minimal or no risk significance) could be subjected to a more refined evaluation to help establish the appropriate regulatory response.

In summary, the FPRSSM process is as follows: (1) a fire scenario is postulated; (2) the inspection findings (fire protection deficiencies) are grouped according to fire area, and then grouped within each fire area according to the particular defense-in-depth element which they impact; (3) the degradation of each defense-in-depth element is characterized qualitatively as severity level high, medium, or low (the severity levels correspond to specific numerical failure probabilities); (4) the failure probabilities are integrated with the ignition frequencies to determine the overall change in CDF for each fire area; (5) using threshold values, the CDF change values are assigned to a performance band (e.g. licensee response band, increased regulatory response band, required regulatory response band). It should be noted that, in this process,.sets of related inspection findings are assessed collectively to determine their synergistic impact on risk. For example, if the plant fire brigade is deficient and the automatic fire suppression system in a fire area is deficient, the FPRSSM considers the adverse impacts of both deficiencies in its assessment of the overall fire risk in the fire area.

Although the FPRSSM was not available during the FPFls or the resulting enforcement proceedings, as a final activity of the FPFI pilot program, the staff has recently used the FPRSSM to assess a sample of the FPFI findings. An overview of the application of this methodology to two sets of inspection findings, one set of high risk significance, and one set of low risk significance, are summarized in Section 10 below.

During a public meeting on March 25, 1999, the staff discussed the proposed FPRSSM with'EI and other interested stakeholders. The staff is working with the reactor oversight task groups to incorporate the FPRSSM into the Inspection Finding Risk Characterization Process described in SECY-99-007A.

10. SAMPLE APPLICATIONS OF FPRSSM Two examples of the application of the proposed Fire Protection Risk Significance Screening Methodology (FPRSSM) are summarized below.

Quad Cities A fire protection inspection was performed at Quad Cities to assess the licensee's corrective actions to address a Confirmatory Action Letter (CAL). These actions included revision and verification of the safe shutdown analysis, development of safe shutdown procedures and identification and resolution of Appendix R discrepancies. The NRC inspection team evaluated a challenging fire scenario involving a fire in turbine building fire area TB-11, a common center section (ground and mezzanine floor elevations) between the Units, 1. and 2 turbines. The area contains switchgear, cables, portions of the turbine lube oil and electro-hydraulic system resenroirs, air compressors, transformers, various electrical panels, and the new resin storage.

This fire area has the potential ignition and fuel sources to produce a fire which is capable of producing a hot gas layer. The inspection team determined that a challenging fire in TB-11 could cause fire damage to redundant trains of safe shutdown functions, require operators of both units to evacuate the main control room and implement a dual unit alternative shutdown which would require manual recovery of safe shutdown equipment and functions for both reactor units.

Using the FPRSSM, the estimated fire initiation frequency (IF) was determined to be on the order of 10 E-2/yr, the degradation in safe shutdown effectiveness (SSD) was judged to be high, the degradation of fire barrier (FB) effectiveness was judged to be high, the degradation in automatic detection/automatic suppression (AD/AS) effectiveness was judged to be medium, and the degradation in detection/manual suppression effectiveness (D/MS) was judged to be medium.

In terms of exponents, the potential risk significance for this scenario and set of identified weaknesses in fire protection defense in depth can be expressed as follows:

i Potential Risk Significance (PRS) = IF+ SSD+ FB+ AD/AS+ D/MS

.. = -2+ . 0 + 0 + -0.75 + -0.5

= -3~ (CDF significantly greater than 10 EA/yr)

This result would prompt further NRC evaluation in order to determine the appropriate regulatory response.

St. Lucie An FPFI was performed at St. Lucie. The inspection focused on St. Lucie Unit 1. The inspection team noted that the licensee had failed to provide, in accordance with a granted exemption, a 1-hour fire rated wrap around those portions of the charging pump 1A cable conduits which extended above the partial height walls separating the redundant charging pumps. Safe shutdown equipment in the charging pump room and an associated 4 foot wide .

passageway common to each pump cubicle include the charging pumps and their associated cables, the CVCS/RWT inter-tie valve and its associated cables, and the pressurizer heater proportional and backup banks control cables. The charging pump room had a low fire loading and no automatic suppression was installed.

0 Using the FPRSSM, the estimated fire initiation frequency was determined to be on the order of 10 E-3/yr, the degradation in safe shutdown effectiveness (SSD) was judged to be low, the degradation in fire barrier effectiveness (FB) was judged to be high, the degradation in automatic detection/automatic suppression effectiveness (AD/AS) was judged to be high, and the degradation in fire detection/manual suppression effectiveness (D/MS) was judged to be medium.

In terms of exponents, the potential risk significance for a charging. pump room fire scenario and set of identified weaknesses in fire protection defense in depth can be expressed as follows:

Potential Risk Significance (PRS) = IF+ SSD+'FB+ AD/AS+ D/MS

=-3+ -3 + 0+ 0 + -0.5

= -6.5 (CDF greater than 10 E-6/yr)

This result would not indicate unacceptable fire protection performance and, in and of itself, would not result in regulatory response above continued NRC "baseline" fire protection oversight.

11. PROGRAM ACCOMPLISHMENTS The following shows how the FPFI pilot program accomplishments, coupled with the proposed baseline fire protection procedures, satisfy the FPFI pilot program objectives.

1. To develop a strong, broad-based, coherent and coordinated NRC fire protection program commensurate with the safety sign Tiicance of the subject.

The proposed fire protection baseline inspection program, combined with availability of the full-scope FPFI procedure when necessary, will provide a much improved and robust NRC fire protection program. The triennial review of licensee fire protection features, configuration management programs, and post-fire safe shutdown capabilities will ensure that the NRC staff is continually well informed of the status of nuclear power plant fire protection.

2. To inspect licensee Thermo-Lag corrective actions and compliance strategies at all plants.

The draft fire protection baseline inspection program will directly review licensee Thermo-Lag corrective actions whether they are based on barrier replacement or re-analysis approaches to Thermo-Lag derating.

3. To inspect non-Thermo-Lag fire protection features in response to ongoing NRC programs (such as self-induced station blackout, fire barrier penetration seals, turbine building assessments, and Individual Plant Examinations of External Events).

Since the draft fire protection baseline inspection procedure will be conducted at each reactor site every three years, the opportunity will exist to rigorously address the site specific programs, procedures and equipment configurations relating to all ongoing NRC programs.

To provide immediate safety benefit through renewed industry attention to nuclear power plant fire safety.

FPFI Workshop industry attendees admitted to the pilot FPFI programs effect of. renewing licensee attention to fire protection programs. The staff believes the draft fire protection baseline insepection program will be equally, if not more effective in doing the same.

Conversely, it is not clear that the recently renewed attention will continue without the existence of frequent and in-depth fire protection inspections.

To develop criteria for licensee fire protection self-assessments.

Workshop participants stated that the draft FPFI procedure (Tl 2515/XXX) is an appropriate basis and technical resource for conducting licensee self-assessments. After the pilot trials of the proposed fire protection baseline inspection procedures, the staff plans to issue the FPFI pilot procedure as a permanent IP. The staff would format the procedure to emphasize its existing modular structure such that individual modules (e.g., fire barriers, fire brigade, safe shutdown capability, etc.) could be performed independently during other types of inspections. The permanent FPFI procedure (1) would be used by the staff to support the triennial inspections as specified in the triennial inspection procedure, (2) would be used by the staff when plant performance falls below a threshold to be established by the new inspection and oversight program, or in response to a specific fire protection event or problem at a plant, and (3) could be used by the licensees as guidance for self-assessments.

To ensure compliance with NRC post-fire'safe shutdown regulations and commitments.

The staff believes that baseline triennial inspections, and reactive use of the FPFI modules that address safe shutdown and configuration management, will ensure compliance with NRC post-fire safe shutdown regulations and commitments through increased licensee attention and corrective actions conducted in response to inspection findings and enforcement actions.

To focus resources. on the fire protection issues of most importance (e.g. licensee control of fire protection design and licensing bases, and those fire protection program elements covered by existing NRC regulations and guidelines).

The staff believes that the draft fire protection baseline inspection program, by virtue of its foundation on FPFI inspection procedure lines of inquiry and inspection techniques, will directly and continuously focus NRC resources on the fire protection issues of most importance.

To evaluate the scope and adequacy of the existing NRC reactor fire protection program, and develop recommendations for program improvement.

The development of the FPFI pilot inspection program was a result of the staff's efforts to evaluate the scope and adequacy of the exisiting NRC reactor fire protection program. The staff believes that implementation of a pilot tested fire protection baseline inspection procedure, combined with selective application of the full-scope FPFI procedure in response to assessments of significantly degraded licensee performance, would constitute a sig'nificant improvement over the previously existing routine fire protection inspection program.

9. To review licensee fire protection and post-fire safe shutdown configuration management.

The FPFI pilot'plant fire protection licensing and design bases was reviewed in detail by the FPFI inspection teams. Fire protection program managment and configuration control is one of the current five major topics addressed in the FPFI procedure (Tl 2515/XXX). Since the proposed triennial inspections use this procedure as a foundation for their inspection planning and conduct, configuration management will receive strong attention if the fire protection baseline inspection procedure is conducted on the regular basis recommended by the staff.

10. To provide clear guidance to the staff and industry regarding oversight of reactor fire protection programs.

\

The staff believes, based on industry feedback as well as FPFI pilot inspection experience, that the FPFI procedure (Tl 2515/XXX) provides clear guidance to the staff and industry regarding oversight of reactor fire protection programs. The staff plans to issue Tl 251 5/XXX as a permanent inspection procedure.

11. To address smoke propagation and manual fire fighting operations, and their impact on equipment operability and operator actions.

During the pilot FPFls smoke propagation, manual fire fighting operations, and their impact on equipment operability and operator actions were addressed in a deliberate manner as directed by the FPFI procedure (Tl 2515/XXX). The inspections included observations of fire drills in risk significant plant areas, and detailed plant tours during which smoke control and manual suppression strategies and capabilities were reviewed.

12. To address balance of plant fire risks.

The inspection reports for the four pilot FPFI inspections, and the Clinton and Quad Cities fire protection team inspections, document the balance of plant inspection results.

13. To improve consistency of internal NRC oversight of licensee fire protection programs.

The staff believes that the proposed baseline monthly resident inspector inspection tours, and the triennial frequency and depth of the proposed baseline team inspections will provide the regions and NRR with a regular source of insightful information regarding the status of licensee fire protection programs. Such of information has been lacking under the existing IP 64704 inspections, especially with respect to post-fire safe shutdown and configuration management program elements.

I

14. To address fire safety considerations not expressly addressed by the fire protection regulation (e.g., event based fires, fire-induced plant transients, seismic/fire interactions,

~

and fire-induced release of radioactive materials).

One inspection finding regarding a potential fire-induced plant transient was found during the River Bend pilot FPFI. This finding involved the possible fire-induced simultaneous opening of all 16 safety relief valves. The related line of inspection inquiry was selected through the review of risk information, including IPEEE results.

APPENDIX A

'IRE PROTECTION FUNCTIONAL INSPECTION PROGRAM ASSESSMENT TREE

APPENDIX B FIRE RISK FACT SHEET

~ The average reported fire frequency at operating plants for the period 1965-1994 is 3.3E-1/yr' The average reported fire frequency for the pre-Appendix R implementation period 1965

- 1985 is 3.8E-1/yr' The average reported fire frequency for the post-Appendix R implementation period 1985 - 1994 is 2.8E-1/yr' During the post-Appendix R implementation period (1986-1994) there were two fire events that resulted in a scram and a loss of function of one safety related division or a loss of 'offsite power. This compares to 10 such events (not including the Browns Ferry fire) during the pre-Appendix R implementation period (1965-1985).'here were 41 fire events that resulted in a plant scram with no loss of function of a safety related division in the 20 year pre-Appendix R implementation period and 40 such events in the 8 year post-Appendix R period.'hirteen large losses from fire events at nuclear power plants during the period from 1966 - 1995 resulted in a total reported monetary loss of approximately $ 800 million, with an average monetary loss per event of approximately $ 62 million.'

On June 21, 1991, the NRC issued GL 88-20, Supplement 4, requesting licerisees to perform an IPEEE to (1) develop an appreciation of severe accident behavior, (2) understand the most likely severe accident sequences, (3) gain a guaiitative understanding of the overall likelihood of core damage and radioactive release, and (4) if necessary, to reduce the overall likelihood of core damage and radioactive release by modifying hardware and procedures that would help prevent or mitigate severe accidents.'pecial Study Fire Events - Feedback of U.S. Operating Experience, June 1997, James R. Houghton, Office for Analysis and Evaluation of Operational Data, USNRC A 30 Year review of Large Losses in the Gas and Electric Utility Industry - 1966- 1995, James B. Biggins, JBH Marsh & McLennan, 1997 NUREG 1407, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," USNRC, June 1991

I Based on the IPEEE results, fire events are important contributors to the reported core damage frequency (CDF) for a majority of plants. The reported CDF contribution from fire events can in some cases, approach (or even exceed) that from internal events."

The reported IPEEE fire CDFs range on the order of E-9/yr to E-4/yr, with the majority of plants reporting a fire CDF in the range from 1E-6/yr to 1EA/yr.'ore than half of the plants proposed or implemented procedural and/or hardware improvements in the fire area in response to their IPEEE.

Although most licensees have reported numerical fire CDF estimates, it is important to note that the accuracy of such estimates has not been validated under the IPEEE submittal review. Because simplifying assumptions and approximate procedures may have been used in the analyses, the quantified CDF estimates reported in the licensees'PEEE submittals should ~onl serve as a general indicator of plant risk. With that in mind the following preliminary information is provided.

a. Fire CDFs for approximately 40 units were greater than or equal to 1E-5/yr.

Fire CDFs for approximately 11 units were greater than or equal to 1E-4/yr.

C. Of those 29 units whose fire CDF was between 1E-5/yr and 1EA/yr, approximately 17 had a reported fire CDF greater than or equal to the reported internal events CDF.

d. Of those 11 units whose fire CDF was greater than or equal to 1EA/yr, 9 units had a fire CDF greater than or equal to the reported internal events CDF. For the remaining 2 units the fire CDF was comparable to the internal events CDF.

January 20, 1998, memorandum to the Commissioners from L Joseph Callan, Executive Director for Operations, Preliminary IPEEE Insights Report 4

This range includes all plants except Quad Cities. The licensee for Quad Cities will submit a revised and updated IPEEE fire analysis during May 1999.

B-2

APPENDIX C

SUMMARY

OF PlLOT FPFI ENFORCEMENT ACTlONS Four pilot inspections were conducted (River Bend, Susquehanna, St. Lucie and Prairie Island.

In addition, during 1998 two.additional inspections were conducted using the draft FPFI procedure. The enforcement actions for all six major fire protection and post-fire safe shutdown team inspections which utilized the draft FPFI procedure are provided below.

1. River Bend Non-cited violation - Lack of a procedure for providing a seismically-qualified source of water to fire protection systems hose stations following a safe-shutdown earthquake.

Discretion exercised to not issue a citation for a Severity Level III violation involving the potential for fire-induced circuit failures in a single multi-conductor cable that could result in simultaneous opening of all (16) safety relief valves, adversely affecting alternative safe shutdown capability.

2. Susquehanna Severity Level IY ( Supplement I) - Indicated reactor vessel water level would not be maintained above the top of the active fuel (TAF) during a postulated Appendix R fire.

Severity Level IV (Supplement I) - Seven instances of fire detection systems or automatic sprinkler, systems not complying with code requirements.

'I Severity Level IV (Supplement I) hour safe shutdown emergency lights not provided. in multiple areas where manual actions were required.

Severity Level IV (Supplement I) - Transient combustible. material in excess of administrative procedures without a permit.

Severity Level IV (Supplement I) - Potential for safe shutdown systems (HPCI, RCIC, CSS, and

.RHR) to be rendered inoperable from discharge piping waterhammer due to unavailability of tools and equipment to install a temporary cross tie hose from the high pressure fire suppression water systems to the condensate transfer system.

Severity Level IV (Supplement I) - Operations Department fire brigade members did not have physical examinations for periods up to two years.

3. St. Lucie Severity Level III, no civil penalty - inadequate alternative shutdown capability procedure to provide adequate guidance to ensure that heating, ventilation and air conditioning equipment to the 1B Electrical Equipment Room and the Hot Shutdown Control Panel Room would be properly operated in the event of a fire in the Control Room or in the Cable Spreading Room.

Severity Level IV, non-cited - Failure to install a one-hour rated fire barrier for conduits carrying cables for charging pump (CP) 1A in accordance with a granted exemption request.

Severity Level IV, non-cited - Incorrect procedural identification of the protected train low pressure safety injection (LPSI) pump.

Discretion exercised to not issue a citation for a Severity Level III violation involving the licensee's failure to analyze for the potential for more than one fire.-induced circuit failure that could cause maloperation of designated safe shutdown equipment.

Discretion exercised to not issue a citation for a Severity Level III violation involving the potential for fire to cause a breach of pressurizer power operated relief valve (PORV) and reactor coolant system gas vent systems (RCSGVS) high/Iow pressure interface boundaries.

Inadequate evaluation of the potential for fire to cause damage to motor operated valves (MOVs) relied upon to accomplish post-fire safe shutdown - final disposition of enforcement is pending.

4. Prairie Island Discretion exercised to not issue a citation for a Severity Level III violation involving the potential for mechanical damage to 32 motor operated valvels (MOVs) resulting from fire-induced circuit failures.

Severity Level IV, non-cited - Failure of fire protection plan to consider the effect of spurious actuation caused by fire damage to equipment needed to support a safe shutdown function.

Severity Level IV, non-cited - Removal of a one-hour rated fire barrier for a credited safe shutdown core cooling pump.

5. Quad Cities Severity Level II, Supplement I - Failure to provide alternative shutdown capability in some fire areas. A postulated fire in certain fire areas would render safe shutdown equipment inoperable such that safe shutdown would not be ensured (14 examples), and a revision to Quad Cities Appendix R procedure was implemented which had not been evaluated and which involved an unreviewed safety question.

6. Clinton Discretion exercised for two violations of Appendix R requirements which had been self-identified during licensee fire protection re'-validation efforts. Both violations involved design control problems, an area for which sign Tiicant NRC enforcement action had previously been taken. One violation pertained to failure to protect motor operated valves from fire induced mechanical damage, and the second violation contained three examples of failure to provide adequate electrical circuit isolation for several safe shutdown components.

C-2

APPENDIX D

SUMMARY

OF STAFF INTERACTIONS WITH THE NUCLEAR ENERGY INSTITUTE On November 10, 1998, the Plant Systems Branch, Division of Systems Safety and Analysis, Office of Nuclear Reactor Regulation (NRR) sponsored a workshop on reactor fire protection inspections in Rockville, Maryland. During that one day conference for public and industry the Nuclear Energy Institute (NEI) proposed a possible course of action consisting of rapid phase-out of FPFls in favor of licensee-managed fire protection program self-assessme'nts. Fire protection discrepancies would be tracked in licensee corrective action systems and result in programmatic changes, procedural changes and/or physical plant modifications. In support of

'his effort, NEI stated that the NRC would develop a number of risk-informed inspection "modules" from the pilot FPFI procedure (Tl 2515/XXX). Over a period of several years all of the modules would be conducted by the licensees. Under the NEI proposal, the NRC would maintain its oversight function through limited self-assessment review inspections. NEI stated that systematic licensee fire protection and post-fire safe shutdown self-assessments should be integrated into the new reactor inspection and oversight program.

During a December 2, 1998 meeting with NEI the NRC staff requested additional information regarding NEI's reactor fire protection initiative, and their expected interfaces with current NRC initiatives. By letter dated January 19, 1999, NEI responded to this staff request. In that letter NEI stated that:

~ The pilot FPFls improved the evaluation of licensee performance, including engineering of plant fire protection systems, compliance with industry standards, passive fire protection features and fire barriers used'to protect safe shutdown functions, post-fire safe shutdown capability, and licensee audits of Appendix R safe shutdown analysis compliance.

However, NRC and industry can achieve these benefits more cost-effectively through appropriate use of FPFI inspection guidance within NEI's proposed industry.FPFI follow-on effort.

~ Licensee self-assessments would be more comprehensive and provide more timely information than continuation of the FPFI program.

~ The results of the FPFls do not appear significant enough to warrant a separate inspection program. The NEI proposal will improve the ability of licensees and the NRC to focus on safety-significant issues.

~ NEI is confident that useful fire protection and post-fire safe shutdown performance indicators can be developed.

In its letter of January 19, 1999, NEI revised its proposal for licensee-managed fire protection program self-assessments in light of SECY-99-007, "Recommendations for Reactor Oversight Process Improvements." This document described risk-informed NRC baseline fire protection inspections which would review ignition sources, control of combustible materials, and fire protection systems and equipment. No performance indicators were identified in this NRC program description. The revised NEI proposal was that:

1. FPFI frequency should be scaled back.

2. Industry and NRC should implement performance indicators in appropriate fire protection and post-fire safe shutdown areas. NRC baseline inspection would be conducted in areas where performance indicators are not available.

3. NRC baseline inspections should utilize "regulatory requirement modules" developed from FPFI inspection guidance. Such "modules" would not include guidance for inspecting potential fire related vulnerabilities as currently specified in Tl 2515/XXX.

4. Formal fire protection self-assessments need not be conducted while licensee performance is in the oversight process "green band," but licensees should conduct progressively more focused self-assessments and should receive increased regulatory attention when performance indicators or baseline inspections indicate a shift in perform'ance to the white, or yellow bands.

5. Licensees should use the NRC developed "regulatory requirement modules" when performing self-assessments.

6. The NRC and licensees should continue other inspections and assessments and credit the results against fire self-assessment activties where appropriate.

7. NRC should conduct team inspections similar to current FPFls only when licensee performance approaches the red "unacceptable performance" band.

The staff met with representatives of the Nuclear Energy Institute (NEI) on March 25, 1999, in Rockville, Maryland to discuss the contents of blEI's January 19, 1999, letter and the new draft baseline fire protection inspection procedure (which was provided to the public and industry the week of March 15, 1999). The NEI representatives stated that the NEI performance indicators issue task force would complete performance indicator pilot trials at selected reactor sites in July 2000. The staff indicated that if NEI develops valid fire protection performance indicators, it would consider their role in the new fire protection baseline inspection procedure at that time.

At the same meeting the NEI representatives stated that an NEI assessment issue task force is developing procedures for voluntary licensee fire protection program self-assessments, with phase-in at commercial reactor sites beginning in summer 1999. The NEI representatives stated that the first stage of the self-assessments would involve plant-specTiic review and risk-prioritization of fire protection commitments, and, depending on the first stage results, follow-on self-assessment activity could be equivalent in scope and depth to that of the NRC's pilot FPFls. The NEI representatives stated that it would be reasonable to expect voluntary self-assessments to be conducted in advance of baseline triennial team inspections, thereby giving the staff an opportunity to reduce inspection burden. The NRC staff noted that the new draft fire protection baseline inspection procedure directs the NRC triennial inspection teams to incorporate the results of licensee self-assessments into their inspection planning.

The new Fire Protection Risk Sign Tiicance Screening Methodology was presented to the NEI representatives at the March 25, 1999 meeting. Examples of the application of this methodology were discussed in detail.

D-2

By letter dated April 14, 1999, NEI submitted comments on the proposed baseline inspection procedures and expressed concerns about including fire protection in the pilot program. The staff will continue to communicate with NEI and other stakeholders as appropriate.

D-3

The Revised R Oversi ht Alan L. Madison Office of Nuclear Reactor Regulation T

Nuclear Regulatory. Commission Washington, D.C.

a we cover...

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Contrasting the old program with the new

'hy Change?

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Key aspects of the new program

~

Future Developments

0 CURRENT OVERSIGHT PROCESS

• Trending Letters

• Superior Performers List

• Watch List SENIOR Other MANAGEMENT Information MEETING PERFORMANCE INDICATORS PLANT SALP Enforcement (AEOD) -

PERFORMANCE (suspended)

REVIEW .

Inspection Program Core Regional Initiative Reactive Generic Safety Issue

Plant Oversight Process C

Management Meeting Action "

Licensee Action Matrix, I

-'- < NRC Inspection, ,

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Regulatory Action I Assessment Report I

I Public Assessment Meeting Enforcement Cornerstones I

Evaluation of Findings Inspection Risk Informed Baseline Per formance

~ Reactive Inspection 'Indicator

~ Initiative Com lementa Su lemenfar Verification

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Strict standards, daily monitoring will continue

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NRC monitoring results easier for public to understand and more readily available

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deregulation, stakeholder input

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U.S. INDUSTRY AVERAGE PERFORMANCE INDICATORS 10

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HOURS 7

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REGULATORY FRAMEWORK NRC's PUBLIC HEALTHAND SAFETY Overall AS A RESULT OF CIVILIAN Safety NUCLEARREACTOR Mission OPERATION REACTOR RADIATION Strategic SAFEGUARD S Performance SAFETY. SAFETY Areas INITIATING MITIGATIO BARRIER EMERGENCY PUBLIC OCCUPATIONA PHYSICAL Corn erstones EVENTS SYSTEMS INTEGRITY PREPAREDNESS PROTECTION HUMAN SAFETY CONSCIOUS WORK PROBLEM -

PERFORMANCE ENVIRONMENT IDENTIFICATIONAND RESOLUTION

" ~ PERFORMANCE INDICATOR

~ INSPECTION

~ OTHER INFORMATIONSOURCES

~ DECISION THRESHOLDS

L Mitigation System O

Hiqh Risk BSC Availability ConlguMion Common Procedure Control Col se Quality Failure Ad use Adequate Maint/Surv Fhrsonnel Rmonnel Controls Operating M@4tein Ftrfowmnce to Maintain to Mantaln ~

Procedure Procedure Rant IAW Doing System Conlg. Plant Design Quality Quality afion Process Accidents Performance

, inspection lnd Inspection Design Ucensed Oper%or Rogrwn Rgb HskSBC Requal Prognwn P AvaIlabiNyindex I

O inspection Pl Validation 8c Vhri5mfion

Table 1- INDIOL'IORS Inabased Regulatory Response Band Initiating Events U~~S ~~pe 7NNQt~H (a~~tcm >3.0 >6.0 >25.0 rrarxal scrams during the pevias fm qLBrters)

Scrams mtha Loss of Normal Heat Rerruvai (over the >4.0 >10.0 >20.0 pevious 12 cgarters)

Unplanned Pcs Orar~ per 7000 Qitical Hours (over >8.0 pevias f(ar qmters)

Mitigating Systems Safety System Unavailability (SSU) AILING+&'ll (average of pevias 12 quarters) Ermrgency Power'll >3.8/o'll Ii

>5.(P/o'll >10.(P/o'll

>2EDG'll >3.$ Yo'll >10.(P/o'll >20.(P/o'll I

HPQ 'll >4.(P/o'll >12.0'/o'll >50.(P/o'll HPCS il >1.PYo'll >4.0//'o li >20.(P/o'll ROC% >4.(P/o'll >12.(P/o'll >50.(P/o'll RHR% >2.(P/o'll >5.(P/o'll >10.(P/o'll Hgh'll HPS1% >2.(P/o'll >5.0/o'll >10.(P/o'll AFVPtl >2.0'/o% >6 (P/o'll >12.(P/o'll RHR >2.(P/o >5.0o/o >10.(P/o Mety System Functional Failures (over pevious >5.0 fan'uarters)

Table I - PERFORMANCE INDICATORS Cont'd Cornerstone indicator Threshoids Increased Required Ur Regulatory Regulatory Pc Response Band Response Band Br Barriers 'H Reac tor Coolant System (RCS) Specific Activity (maximum >50.0% >100%

~Fuel Cladding 'll monthly values, percent of Tech. Spec limit, during revious four quarters)

~Reactor Coolant RCS Identified Leak Rate (maximum monthly values, >50.0% > 100%

System 'll percent of Tech. Spec. limit, during previous four quarters)

~Containment Containment Leakage (maximum monthly values, >60.0% N/A ercenta e of LA over the revious'our quarters)

Emergency Drill/Exercise Performance (over previous eight quarters) <90.0% <70.0%

Preparedness ERO Drill Participation (percentage of Key ERO personnel <80.0% 'll <60.0% 'll that have participated in a drill or exercise in the previous ei ht quarters)

Alert and Notification System Reliability (percentage <94.0% <90.0%

reliability during previous four quarters)

Occupational 'll Occupational Exposure Control Effectiveness (occurrences >5 N/

Radiation Safety durin revious 12 uarters)

Public RadiatIon RETS/ODCM Radiological Effluent Occurrence N/

Safety (occurrences durin revious four uarters)

PhysIcal Protection Protected Area Security Equipment Performance Index >0.05 >0.15 (over a four quarter eriod)

Personnel Screening Program Performance (reportable >2 N/

events durin the revious four uarters)

Fitness-for-Duty (FFD)/Personnel Reliability Program N/

Performance (reportable events during the previous four

Performance Indicators Action Impact Matrix on the Licensee Inspection Findings

PERFORMANCE t

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ALUATINGLICENSEE PERFORMANCE INDICATORS:

CONCEPTUAL MODEL GREEN Licensee Response Band Cornerstone objectives f'ully met. Nominal risk with; normal deviation from expected performance.

WHITE Increased Regulatory Response Band Cornerstone objectives met with minimal reduction in safety margin.

Changes in performance consistent with 6 CDF( E-S (h LERF( E-6)

YELLOW Required Regulatory Response Band Cornerstone objectives met with significant reduction in safety margin.

Changes in performance consistent with h, CDF( E-4 ( 6, LERF( E-5)

RKD P/ants not permitted to operate within this Band Plant performance significantly outside design basis. Loss of confidence in ability of plant to provide assurance of public health and safety with continued operation. Unacceptable margin of safety.

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ACTION MATRIX IN REA IN LICENSEE PERFORMANCE AFETY I NIFI AN E ->

V) All Assessment One or Two White Inputs. One Degraded Repetitive Degraded Overall Unacceptable I- inputs (Performance (in different Cornerstone (2 White Cornerstone, Multiple Performance; Plants Not Indicators (Pls) and cornerstones) in a Inputs or 1 Yellow Input) Degraded Cornerstones. Permitted to Operate (0 Inspection Findings) Strategic Performance or any 3 While Inputs in Multiple Yellow Inputs, Within this Band, tO Green; Cornerstone Area; Cornerstone a Strategic Performance or 1 Red Input';

IL' Unacceptable Margin to Objectives Fully Met Objectives Fully Met Area; Cornerstone Cornerstone Objectives Safety Objectives Met with Met with Longstanding Minimal Reduclion in Issues or Significant Safety Margin Reduction in Safety Margin Regulatory Rouline Senior Branch Chief (BC) or DD or Regional EDO (or Commission) Commission meeting with Conference Resident Inspector Division Director (DD) Administrator (RA) Meet Meet with Senior Senior Licensee (SRI) interaction Meet with Licensee with Licensee Licensee Management Management Licensee Action Licensee Corrective Licensee Corrective Licensee Self Licensee Performance Action Action with NRC Assessment with NRC Improvement Plan with Oversight Oversight NRC Oversight NRC Inspection Risk-Informed Baseline and Inspection Baseline and Inspection Baseline and Team 0 Baseline Inspection Follow-up Focused on Cause of Inspection Focused on Program (Baseline) Degradation Cause of Degradation Regulatory None Document Response to Docket Response to -10 CFR 2.204 DFI Order to Modify, Suspend, Actions Degrading Area in Degrading Condition -10 CFR 50.54(f) Letter or Revoke Licensed n i R AVr r I Assessment DD review/sign DD review/sign RA review/sign RA review/sign O Report assessment report assessment report assessment report assessment report (w/inspection plan) (w/ Inspection plan) (w/ inspection plan) (w/ inspection plan)

O Commission Informed Public SRI or BC Meet with BC or DD Meet with RA Discuss EDO (or Commission) Commission Meeting with o Assessment Licensee Licensee Performance with Discuss Performance Senior Licensee Meeting Licensee with Senior Licensee anagement Management

<-----R i n IR vi w IA enc Review -

REVIEW SYSTEM .

t.evel of Frequeneyf I'artlolpant ~ Desired Communtcatlon Revtevr Tlmlng ('ndlcatea load) Outcome Continuous Contlnueua IAI', Al, regional Performance Inspectors, analysts awareness Once per quarl ttl OAP: BC',Ptl, fnputfverlfy PfIPN Updated data set Ywo weeks after SRl, Rl da'la detecl

~ nd of quarter ~ arly trends ht mtdwycfer Olvlslons of Reactor, Detect trends, Slx month Three wee)is Safety (OAS) or ORP plan Inspection Inspection after end of 00', ORP <<nd DRI for six months took ahead tetNr second quarter Bca RnbofCycle At enbelwyefef . DAS or DRP DD', Assessment Assessment fetter Four weeks AAs, NRR of plant and six month

~ ftef end of representative, BCs, performance, lnspecllon foott assessment prlnclpal inspectors, epproveI ahead fetter cycle OE,OI, olher HQ coordinate offices es approprlke regional actions Agency Annuattyf DlR NRR; RAs, ApproM Commlsslon Action 'eo weeks DRSIDRP DDe, coordinate brfeflng, followed Aevtew aflef end+- AEOD, DISP, O'E, Df, agency by public meellnga cycle levtesr other HQ offices as actions with lndlvldual a ppropfla'te licensees to discuss assessment results

&ctotlms snt senior Raskfent Inspector DD Dfvts ton Director Al Reekfent Inepector RA tleglonal Admlntatratet BC Branch Chief DIR Director PE Pro}oct Kntffneer DNP Dlvfston of Inspectton and Ouptxxt Prolrarna DRP Dtvtslon ot Reactor pie)acts . Ol ONce of Invastltfatlone

hange to our Enforcement Policy

~

NEW POLICY Continue current policy for willfulviolations, radiation over exposures, releases of radioactive material, failure to report required information to XRC Other types of violations a Regulatory Conference can be held to discuss significance. Notice of Violations will continue to be issued based on safety significance Civil Penalties willnot normally be imposed. The new process better determines NRC response.

H THK PUBLI BK KFIT

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More information, more often, more readily available, more understandable

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Predictable and consistent actions by the regulator based on plant performance

~ Focus on more risk-significant issues enhancing safety

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PERFORMANCE INDICATOR

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Fire Protection Indicators

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Unreliability Indicators

~ Risk-based Indicators

~ Maintain safety by establishing regulatory oversight framework that ensures continued safe operation

~

Enhance public confidence

~

Improve effectiveness and efficiency of oversight process by focusing resources on most risk significance

~ Reduce unnecessary regulatory burden

FIRE PROTECTION RISK SIGNIFICANCE SCREENING METHODOLOGY Probabilistic Safety Assessment. Branch Plant Systems Branch Office of Nuclear Reactor Regulation, NRC (Presented at Region IV PRA Workshop; July 20,$ 999)

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~

OB JECTIVES

~ Focuses resources on monitoring performance and effectiveness of fire protection mitigation features important to risk

~ Provides a two-phase method for characterizing the.

risk significance of inspection findings - screens out findings with minimal or no risk significance

~ Fits into plant oversight assessment process, thus recommending a regulatory response due to the potential risk significance of fire protection inspection findings

i: (

OVERVIEW Phase 1

~ Screen out individual inspection findings not affecting DID

~ Primary user is resident inspector

~ Does not require fire protection DID to be evaluated fully for fire area Phase 2

~ To be'entered if finding(s) do not screen in Phase 1, OR during fire protection triennial inspection

~ Primary user is Region

~ Evaluate DID fully for fire area to assess hCDF.

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PHASE 1 Two-Step Process for Phase 1 Screening

~ Step 1: Impact on function of DID, relationship to AOT

. ~ Step 2: SSD for the fire area, fire protection scheme

~ All equipment and cables in fire area are assumed failed

~ If DID principle not evaluated, it is assumed to have a low degradation

PHASE 2

~ Is Conservative. If a fire scenario can be developed, then

~ . All equipment and cables in room where fire initiates is failed. The barrier is challenged, and if failed, all equipment and cables in adjacent room are failed too.

~ Characterization of DID degradation due to inspection findings is conservative.

~ Is Qualitative

~ Degradations in DID are categorized as High, Medium, or Low

PHASE 2 (cont.}

~ Views inspection findings collectively

~ Synergistic impact on risk of a fire area

~ Feeds into "Inspection Finding Risk Characterization Process" (SECY-99-007A, attachment 2)

~ Fire mitigation frequency (IF, AS, MS, FB) integrated with SSD mitigation capability from SDP tool to approximate 8, CDF

~ Therefore, produces risk significance categories which are consistent with regulatory response thresholds used in NRC licensee performance assessment process.

UNDERLYING QUANTITATIVEFOUNDATION

~ Failure probabilities developed for qualitative degradations in DID

~ Product of ignition frequency and failure probabilities for DID produces CDF

~ Dependencies between auto-suppression and manual suppression modeled r

~ Since auto-suppression only controls fire, credit for auto-suppression adjusted with fire brigade high degradation

~ Common mode failures of water based auto-suppression and fire brigade included

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Fire Protection Risk Significance Screening Methodology Example of its Application

TRANSFORMER VAULTROOM Room protection Hour Rated Room (doors/dampersipenetration seals)

Startup and Unit Auxiliary Transformers Housed in the Room Redundant Post-Fire Safe Shutdown Located in the Room SW Pumps (power and control cables)

CCW Pumps (power and control cables) 125 Ydc Power Distribution Cables Fire Protection Oil Containment Curb Around Transformers Automatic Water Spray System Actuated by Cross Zone Fire Detection Provided For Transformers 1-Hour Fire barrier Provided for One Train of Post Fire Safe Shutdown Functions

Screenin Process Phase1 Ste 1 Fi ure4 For a given fire area, zone, or room under consideration Degradation or Clearly stated fire No impairment or impairment of DID protection degradation of fire element was less than findings protection feature or the allowed outage time DID effectiveness without the appropriate compensatory measure.

Or Affects one of the following Yes, Screen Degradatlon or

=fire mitigation DlD elements: impairment of D1D element existed for less than 30 days with the

1. Detection and manual suppression capability appropriate compensatory measure

2. Automatic suppression capability No, screen

3. Fire barriers Yes, Yes, Screen Go to step 2 of Phase f

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0 Fire Protection Schemes (Appendix R of 10 CFR Part 50, Section III.G.2)

Scheme 1 Provide a 3-hour fire barrier separation which either encloses one SSD train or provides wall to wall and floor to floor separation between the redundant trains; or Scheme 2 Provide a 0-hour fire barrier enclosing one of the SSD trains. The area shall be protected by automatic fire detection and suppression; or Scheme 3 Provide greater than 20 feet of horizontal separation between the redundant SSD trains. The spacial separation between the two trains shall be free of intervening combustibles. The area shall be protected by automatic fire detection and suppression.

Yes No ls the automatic ls Protection ls 1-hour fire barrier fire suppression Scheme 2 used? that encloses one SSD system affected No, function affected by by the finding?

Screen finding?

Yes Yes ls recovery system physically independent (separated by a 3-hour fire barrier) of the fire area, zone, or room of concern and capable of being manually actuated under the time constraints?

No, Yes, perform Phase 2 Screen

FPRSSM - Phase 2 Methodology Steps Grouping of findings Define fire scenario Qualitative evaluation of findings Assignment of quantitative values Determination of fire ignition frequency integrated assessment of DID findings and fire ignition frequency Integration of adjusted Fire Mitigation Factor (FMF) with safe shutdown General rules for applying FPRSSM

INSPECT'ION FINDINGS (Step 1) (Example 1)

VfATER SPRAY SYSTEM Re-assembly after startup transformer modifications Relocate spray nozzles so that they were not aimed at the hazard tape'from prior. system painting found covering the orifice I'asking openings on some nozzles 3 of 6 rate compensated fire detectors on the startup transformer had their outer shell (metal tube) dented. The tubes had through wall cracks (possibly affecting the outer shell's expansion coefficient)

ELEGTRIGAL RACEWAY FIRE BARRIER SYSTEM 1-hour Fire Barrier Conduit Protection on Sw Cables Has Through Barrier Opening MANUALFIRE FIGHTING EQUIPMENT No Apparent Problems with the Hose Station Equipment or Extinguishers

FIRE BRIGADE DRILL OBSERVATIONS No radio communications (used cell phones - phones did not work)

Hose deployment problems, hose not long enough to reach the fire Turnout protective clothing not properly protecting personnel Use of personal protective equipment not adequate (didn't use SCBA, clothing not donned properly)

Fire attack techniques were not proper, pre-fire plans did not have a smoke control plan Did not bring proper equipment

TULATED FIRE SCENARIO (Step 2)

Fire Likelihood Not Assessed Define a Fire Condition in the Room That ls Capable of Developing a Hot Gas Layer or a Direct Exposure to Critical Systems, Equipment, or Components (Bigni[icant Fire)

A Possible Fire Conditiori Fault offsite cause a fault and failure of the startup transformer Transformer housing fails and releases burning oil post-fire Safe Shutdown - Possible lm act Loss of off site power (due to loss of switchyard)

Loss of Service Water to the EDGs and CCW 10

EXAMPLE 1

~ FB (1 hour} = High, AD/AS = High, D/MS = High ($

~+p83and4)

~ IF ) 1E-3/QI'step 5)

~ Fire Mitigation Factor (FMF) = IF+ FB+ MS +AS+CC (when appropriate}

Where IF = Fire Ignition Frequency FB = Fire Barrier MS = Manual Suppression/Detection AS = Automatic Suppression/Detection CC = Dependencies/Common Cause Contribution

~ FMF = IF + FB + AD/AS + D/MS (Step 6)

-2+ 0 + 0 + -0.25 = -2.25

~ -2 > FMF >-3 (1 per10'to10')

~ Condition greater than 30 days (estimated likelihood rating C)

~ Risk significance estimation = RED (step 7) 11

-Table 5.6 Association of FMF to Table.5.?

(SDP Table 1j approximate frequencies for Calculation of'Delta CDF Fire Mitigation Frequency (FMF) Table 5.7- Approximate Frequencies FMF >-2 1 per10to10'

-2 o'MF )-3 per10'to10'

-3 ~e FMF >-4 per 10'o 104

-4 > F.MF. >-5 1 per104to10'

-5 o'-MF )-'6 per10'to10'ess e

FMF <--6 than 10'2

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INSPECTION FINDINGS (Example 2)

No Apparent Problems with the Water Spray System 1-hour Fire Barrier Protecting Post Fire Safe Shutdown Functions had mechanical damage that reduced the barrier's wall thickness by 30%

No Apparent Problems With the Hose Station Equipment or Extinguishers Fire Brigade {effectiveness and efficiency) Drill Observations Poor radio communications (Radios did not work properly)

Hose deployment problems Turnout protective clothing not properly protecting personnel Minor fire attack techniques problems 15

EXAMPLE 2

~ FB (1 hour} = Medium, AD/AS = Low, D/MS = Medium (steps 3and4i

~ IF >1E-3/yr (step5}

~ FMF = IF+ FB + AD/AS+ D(MS (step6)

-2 + -0.5 + -1.5 + -0.5 = -4.5

~ -4) FMF ) -5 (1 per 10'to10'}

~ Condition greater than 30 days (estimated likelihood rating E}

~ Risk significance estimation = WHITE (pep 7) 16

INSPECTION FINDINGS (Example 3}

No Apparent Problems with the Water Spray System No Apparent Problems with the Fire Barrier Systems Protecting Safe Shutdown Functions No Apparent Problems Vfith the Hose Station Equipment or Extinguishers Fire Brigade (effectiveness and efficiency} Drill Observations Poor radio communications (Radios did not work properly}

Hose deployment problems Turnout protective clothing not properly protecting personnel Minor fire attack techniques problems

E AMPLE3

~ FB {1 hour} = Low, AD/AS = Low, D/MS '= Medium isteps p and 4)

~ IF ) 1E-3/pl" (step 5)

~ FMF = IF + FB + ADIAS + D/MS {step 6)

-2+ -1 + -1.5 + -0.5 = -5.0

~ -6 ) FMF) -6 {1 per 10'to 10'}

~ Condition greater than 30 days {estimated likelihood rating F} .

0 Risk significance estimation = GREEN (step 7) 18

SUMMARY

,Exam le1

~ FB (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)= High, AD/AS = High, D/MS = High

~ Potential Risk Significance (with recovery of a train) = Red (without recovery of a train) = Red Exam le2

~ FB (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) = Medium, AD/AS = Low, D/MS = Medium

~ Potential Risk Significance (with recovery of a train) = White (without recovery of a train) = Yellow Exam le3

~ FB (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) =Low, AD/AS=Low, D/MS=Medium

~ Potential Risk Significance (with recovery of a train)= Green (without recovery of a train) = White

EXAMPLE t TRANSFORMER VAULTAREA PHASE 2 FPRSSM WORKSHEET FOR FIRE-INDUCED LOOP

~ WITHOUT RECOVERY

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11 L ORVCR - LPR 12 LO - PORVCR - HPR 13 LOOP - PORVC - CH - FB 14 LOOP - PORVC - CH - Sl Identify any operator recovery actions'hat are credited to directly restore the degraded equipment:

Note 1: If operator actions are required to credit placing mitigation equipment in service or for recovery actions, such credit shoutd be given only if the following criteria are met: 1) sufficient time is available to implement these actions, 2) environmental conditions allow access where needed, 3) procedures exist, 4) training is conducted on the existing procedures under conditions similar to the scenario assumed, and

5) any equipment needed to complete these actions is available

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DRT/SRT,DRT Fire Area/Zone TRANS VAULTAREA FMF (Table 1 Row) -2.25 Table i result(circle): A B C D E F G H Safet Functions Needed: Full Creditable Miti ation Ca abilit for each Safet Function:

Emergency AC Power (EAC) 2/3 Emergency Diesel Generators (3 EDGs= 1 multi-train system, 2EDG=1 diverse train) or 1 Gas Turbine Generator (1 diverse train)

Recovery of AC power in < 6 hrs (REC6) Recover a source of AC to allow primary injection (Operator action under high stress)

(

Recovery of AC Power in 2 hrs (REC2) 1 TDAFW train and SBO procedures, other than GTG, implemented (operator action under high stress)

Early Inventory, HP Injection (EIHP), 1 / 2 Charging trains (1 multi-train system) or 1 /2 Sl trains (1 multi-train system)

Secondary Heat Removal (AFW) 1 TDAFW train (1 train) or 1 /2 MDAFWtrains (1 multi-train system)

Primary Heat Removal, Feed/Bleed (FB) 2 /2 PORVs open for Feed/Bleed (operator action under high stress)

Low Pressure Recirc (LPR) 1 /2 LPSI trains (1 multi-train system) .

High Pressure Recirc (HPR) 1 /2 Charging trains (CH) or 1/2 SI trains taking suction from 1 /2 LPSI trains (1 multi-train system but also requires human action for switching the suction to sump)

PORVCR PORVs challenged and fail to reseat (-2 for random)

PORVC PORVs challenged (-1 for random)

Circle affected functions Recoveror of S~urioue Remainin Miti ation Ca abilit Sum ~Se uence failed train Actuation Color 1 LOOP - EAC - REC6 REC6 (-1) Red 2 LOOP - EAC - REC2 - TDAFW REC2 (-1) TDAFW (-2) 3 LOOP - EAC - EIHP (RCP seal LOCA) EIHP (-6) 4 LOOP - EAC - REC2 - FB (RCP seal LOCA) Can see that the remainder will be substantially 5 LOOP - EAC - REC2 - LPR (RCP seal LOCA) less than-1. Therefore-1 isthesum. (Note 6 LOOP - EAC - REC2 - HPR (RCP seal LOCA) that also with a Red you could stop anyhow) 7 LOOP - AFW- FB 8 LOOP -AFW- LPR 9 LOOP-AFW- HPR 10 LOOP - PORVCR - EIHP

11 LO ORVCR - LPR 12 LO PORVCR - HPR 13 LOOP - PORVC - CH - FB 14 LOOP - PORVC - CH - SI Identify any operator recovery actions'hat are credited to directly restore the degraded equipment:

Note 1: If operator actions are required to credit placing mitigation equipment in service or for recovery actions, such credit should be given only if the following criteria are met: t) sufficient time is available to implement these actions, 2) environmental conditions allow access where needed, 3) procedures exist, 4) training is conducted on the existing procedures under conditions similar to the scenario assumed, and

5) any equipment needed to complete these actions is available Note that SRT for each of these will be significantly less due to EDGs and other random failures necessary for core damage. Therefore, only need DRT for any condition of 1 hr. barrier (high, medium, or low)

Attachment 4 ANSWERS TO QUESTIONS RAISED DURING THE FIRE PROTECTION BASELINE PILOT INSPECTION LICENSEE PRE-BRIEF MEETING Question: In the New Reactor Oversight and Assessment Process it appears that, multiple inspection findings of the same degree of importance can be summed to represent a finding of the next higher degree of importance. For example, findings which are individually risk categorized as "Green - Licensee Response Band," can cumulatively result in a White - Increased Regulatory Response Band" status for.a giv'en licensee performance "cornerstone." Is this true, and what criteria would be applied to raise a group of findings from one band to the next?

Answer: When related findings occur simultaneously, their risk significance will be evaluated collectively within the Significance Determination Process (SDP). Each findin or set of simultaneous findin s evaluated by the SDP may affect more than one core damage sequence. If the change in likelihood of multiple sequences includes three or more "green" sequences which are just below a "white" characterization (i.e., each less than 1E-6/yr but greater than or equal to 1E-7/yr), then current SDP guidance requires considering the sum of these sequence likelihood changes as increasing overall core damage frequency (CDF) enough to result in a "white" (greater than

-6/yi 1 g f ~hl dpi In general, the changes to core damage frequency (CDF) from independent, unrelated findings identified during an inspection, each of which has been individually characterized as "green" within the SDP process, are not summed to "white" in the new reactor oversight and assessment process. However, a single fire brigade finding may pertain to multiple fire areas, and the associated "green" CDF changes for various fire areas may combine to produce a "white" plant indication. Guidance for addressing this situation is under consideration.

Question: Aren't NRC granted exemptions generally risk significant, and therefore likely to be identified as risk significant within the Fire Protection Risk Significance Screening Methodology (FPRSSM) which will be applied during the fire protection baseline pilot inspections?

Answer: No. SECY-99-182 "Assessment of the Impact of Appendix R Fire Protection Exemptions on Fire Risk", dated July 9, 1999, described a review by Sandia National Laboratory of 169 fire protection related exemptions from a sample of 9 reactor plants (1 3 units). Only five were identified as potentially risk significant in accordance with the criteria contained in Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-informed Decisions on Plant-specific Changes to the Licensing Basis'July, 1998). Eighty percent of the exemptions reviewed were:found to have a very small risk impact.

Question: CCDPs are the result of assuming that 1) a fire occurs and 2) the fire damages all equipment in the area, or the fire damages a specific subset of cables and equipment in the area based on fire modeling or separation assumptions. Are the fire protection baseline inspection teams going to focus on areas of the reactor plants with high core damage frequencies (CDFs), or also be looking at areas with high conditional core damage probabilities (CCDPs).

Answer: The fire protection baseline inspection teams do not intend to focus exclusively on plant areas with high CDFs. The licensee's analyses and assumptions which establish that plant areas with high CCDPs are actually areas with low CDF may be rev'iewed. The team will either have access to licensee developed CCDP information for the subject reactor plant, or the regional Senior Reactor Analyst (SRA) will generate the needed CCDP values.

Question: Where are the FPRSSM evaluation criteria for fire brigade effectiveness?

Answer: In Section 9.1 of Appendix H, "Guidance for Making a Qualitative Assessment of Fire Protection Inspection Findings - Fire Protection Risk Significance Screening Methodology," of the fire protection supplemental inspection procedure (FPSI). This procedure will be issued before the pilot fire protection baseline inspections begin.

Question: What are the guiding documents for the pilot inspections?

Answer: Attachment 05 of IP 71111 is the fire protection baseline inspection procedure.

Appendix H of the fire protection supplemental inspection (FPSI) procedure will provide fire protection degradation evaluation criteria to support application of the Fire Protection Significance Screening Methodology (FPRSSM), which is provided in the new 'isk reactor oversight and assessment process's manual chapter titled."Significance Determination Process (SDP)."

Question: What versions of the NFPA codes will the triennial fire protection inspection teams and the resident inspectors use to assess fire protection features and degradations?

Answer: The inspectors will determine whether or not fire protection features meet the NFPA codes of record for the licensed reactor facility. However, the as-installed configuration and material condition of the installed fire protection equipment and features, the physical plant arrangement relative to the fire protection equipment and features, and licensee performance issues (e.g. fire brigade capabilities, control of ignition sources and transient combustibles) will be considered in the development of fire protection findings. Fire protection findings will, in turn, be entered into the Fire Protection Risk Significance Screening Methodology (FPRSSM) of the Significance Determination Process (SDP).

Attachment 5 Mr., President Licensee Nuclear Department Licensee Corporation or Company Address

SUBJECT:

SELECTED NUCLEAR POWER STATION, UNITS 1 AND 2 - NOTIFICATiON OF CONDUCT OF A TRIENNIALFIRE PROTECTION BASELINE INSPECTION

Dear Mr.:

The purpose of this letter is to notify you that the U.S. Nuclear Regulatory Commission (NRC)

¹ Region staff will conduct a triennial fire protection baseline inspection at Selected Nuclear Power Station, Units 1 and 2 in Month, 20¹¹. The inspection team will be lead by Mr. First Last,

¹ from the NRC Region Office. The team will be composed of personnel from NRC Region ¹,

and Contracted National Laboratory. The inspection will be conducted in accordance with IP 71111.05, the NRC's baseline fire protection inspection procedure.

The schedule for the inspection is as follows:

Information gathering visit - Month ¹¹-¹¹, 20¹¹ [Note - this date is pre-coordinated with the licensee]

Week of onsite inspection - Month ¹¹, 20¹¹.

In advance of the first week of onsite inspection, members of the inspection team will visit Selected Nuclear Power Station, Units 1 and 2, for two days during the week of Month ¹¹, 20¹¹ to obtain information and documentation needed to support the inspection, to become familiar with the Selected Nuclear Power Station, Units 1 and 2 fire protection programs, fire protection features, post-fire safe shutdown capabilities and plant layout, and, as necessary, obtain plant specific site access training and badging for unescorted site access. A non-exhaustive list of the types of documents the team. will be interested in reviewing, and possibly obtaining, are listed in Enclosure 2.

During the information gathering visit, the team will coordinate with your staff on, the following inspection support administrative details: office space size and location; specific documents requested to be made available to the team in their office spaces; arrangements for reactor site access (including radiation protection training, security, safety and fitness for duty requirements); and the availability of knowledgeable plant engineering and licensing organization personnel to serve as points of contact during the inspection.

We request that during the onsite inspection week you ensure that copies of analyses, evaluations or documentation regarding the implementation and maintenance of your fire protection program, including post-fire safe shutdown capability analyses, be readily accessible to the team for their review. Of specific interest are those documents which establish that your fire protection program satisfies NRC regulatory requirements, conforms to applicable NRC fire protection guidance, and industry fire protection codes and standards. Also, appropriate personnel, knowledgeable with respect to those plant systems required to achieve and maintain safe shutdown conditions from inside and outside the control room (including the analyses supporting the post-fire safe shutdown methodology), reactor plant fire protection systems, and

the Selected Nuclear Power Station fire protection program implementation should be available at the site during the inspection.

Your cooperation and support during this inspection will be appreciated. If you have questions concerning this inspection, or the inspection team's information or logistical needs, please

¹ contact First Last, the team leader, in the Region Office at ¹¹¹A¹¹C¹¹¹.

Sincerely, Docket Nos.: 50-¹¹¹ and 50-¹¹¹

Enclosure:

As stated (1)

ENCLOSURE 1

, Reactor Fire Protection Pro ram Su ortin Documentation The current version of the Fire Protection Program and Fire Hazards Analysis.

Current versions of the fire protection program implementing procedures (e.g.,

administrative controls, surveillance, testing, fire brigade).

Fire brigade training program and pre-fire plans.

4. Post-fire safe shutdown analysis and related analyses which demonstrate that fire-induced circuit failures will not result in spurious signals or uncontrolled equipment operations which may prevent required post-fire safe shutdown equipment from performing their intended functions during and after a postulated fire.

Plant layout and equipment drawings which identify the physical plant locations of hot standby and cold shutdown equipment.

6. Piping and instrumentation (flow) diagrams highlighting the preferred flowpaths and components needed to achieve and maintain hot standby and cold shutdown for the various fire areas outside the main control room and for those systems and components used for plant areas relying on alternative shutdown capability.

Plant layout drawings which identify plant fire area delineation, areas protected by automatic fire suppression and detection, and the locations of fire protection equipment.

8. Plant layout drawings which identify the general location'of the post-fire emergency lighting units.

Plant operating procedures which would be used and describe shutdown from inside the control room with a postulated fire occurring in any plant area outside the control room, procedures which would be used to implement alternative shutdown capability in the event of a fire in either the control or cable spreading room.

10. Maintenance and surveillance testing procedures for alternative shutdown capability, fire barriers, detectors, pumps and suppression systems.

Maintenance procedures which routinely verify fuse breaker coordination in accordance with the post-fire safe shutdown coordination analysis.

12. The reactor plant's IPE and IPEEE.

13. Organization charts of site personnel down to the level of fire protection staff personnel.

14. If applicable, drawings indicating potential reactor coolant/recirculation pump lube oil system leakage points, and the oil collection system provided for those potential leakage points.

0

15. The SERs and 50.59 reviews which form the licensing basis for the reactor plant's post-fire safe shutdown configuration.

16. A list of applicable codes (e.g., NFPA codes of record) and standards related to the design of plant fire protection features and evaluations of code deviations.

17. Recent fire protection QA audits and/or fire protection self-assessments.

18. Recent QA surveillances of fire protection activities.

19. Listing of open,and closed fire protection condition reports (problem reports/NCRs/EARs/problem identification and resolution reports).

20. Listing of Generic Letter 86-10.evaluations.

0