Text

Proposed Radiological Effluent Tech Specs
	ML20101N381
Person / Time
Site:	FitzPatrick
Issue date:	12/21/1984
From:	; POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	;
Shared Package
ML20101N373	List: ML20101N370; ML20101N375; ML20101N381;
References
	NUDOCS 8501040006
	Download: ML20101N381 (103)
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i ATTACHMENT I to JPN 86 l

! PROPOSED RADIOLOGICAL EFFLUENT i.

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! TECHNICAL SPECIFICATIONS (RETS) l-l (JPTS-83-09) l l

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- NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 f DPR-59 l

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l RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS APPENDIX B TO FACILITY OPERATING LICENSE NO. DPR-59 FOR JAMES A. FITZPATRICK NUCLEAR POWER PLANT i

NEW YORK POWER AUTHORITY

(- DOCKET NO. 50-333 l

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. Amendment No.

RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS Table of Contents

2. age 1

1.0 Definitions 2.0 Liquid Effluents 3 2.1 Liquid Effluent Monitors 3 2.2 Concentration of Liquid Effluents 6 2.3 Dose from Liquid Effluents 10 2.4 Liquid Radioactive Waste Treatment System Operations 12 2.5 Maximum Activity in Outside Tanks 14 3.0 Gnceous Effluents 17 3.1 Gaseous Effluent Monitors 17 3.2 Gaseous Dose Rater 19 3.3 Air Dose, Noble Gases 24 3.4 Dose Due to Iodine-131, Iodine-133. Tritium and Radionuclides in Particulate Form 26 3.5 Main Condenser Steam Jet Air Ejectors (SJAE) 28 3.6 Offgas Treatment System 30 3.7 Offgas Treatment System Explosive Gas Mixture Instrumentation 32 3.8 Standby Gas Treatment System (SBGTS) 34 3.9 Mechanical Vacuum Pump Isolation 35 3.10 Main Control Room Ventilation Radiation Monitor 36 l

4.0 Solid Radioactive Waste 43 4.1 Process Control Program 43 5.0 Total Dose 46 i 5.1 Total Dose from Uranium Fuel Cycle 46 6.0 Radiological Environmental Monitoring 51 6.1 Monitoring Program 51 6.2 Land Use Census Program 61 6.3 Interlaboratory Comparison Program 63 7.0 Administrative Controls 66 7.1 Responsibility 66 7.2 Procedures 66 7.3 Reporting Requirements 66 Amendment No. i

RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS List of Tables Table Title Page 2.1-1 Radioactive Liquid Effluent Monitoring Instrumentation 5

2.2-1 Radioactive Liquid Waste Sampling and Analysis Program 7 3.2-1 Radioactive Gaseous Waste Sampling and Analysis Program 21 l

3.10-1 Radiation Monitoring Systems that Initiate and/or l Isolate Systems 37 3.10-2 Minimum Test and Calibration Frequency for Radiation Monitoring Systems 38 6.1-1 Operational Radiological Environmental Monitoring Program 54 l 6.1-2 Reporting Level for Radioactivity Concentrations in Environmental Samples 58 6.1-3 Detection Capabilities for Environmental Sample Analysis 59 s

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l Amendment No. ii

RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS List of Figures Figure Title g 5.1-1 Site Boundary Map 48 l

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Amendment No. iii

RADIOLOGICAL EFFLUENT TECHNICAL SPEC FICATIONS 1.0 DEFINITIONS A. Dose Equivalent I-131 The Dose Equivalent 1-131 is the concentration of I-131 (nierceu-rie/ gram) which alone would produce the same thyroid dose ao 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 III of TID-14844, "Caled-lation of Distance Factors for Power and Test Reactor Sites" or in NRC Regulatory Guide 1.109 Revision 1, October 1977.

B. Instrument Channel Calibration See Appendix A Technical Specifications.

C. Instrument Channel Functional Test See Appendix A Technical Specifications.

D. Instrument Check See Appendix A Technical Specifications.

E. Logic System Function Test See Appendix A Technical Specifications.

F. Member (s) of the Public Member (s) of the Public includes all persons who are not occupation-ally associated with the facilities on the NYPA/(NMPC) Niagara Mohawk l Power Corporation site. This category does not faclude employees of l the utilities, its contractors or vendors. Also excluded from this l 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 plants, i G. Offgas Treatment System The Offgas Treatment System is the system designed and installed to:

reduce re aa'.tive gaseous effluents by collecting primary coolant system offg:ses from the main condenser; and, providing for delay of '-

the offgas for the purpose of reducing the total radioactivity prior l

to release to the environment.

H. Offsite Dose Calculation Manual (ODCM)

The ODCM describes the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid .

effluents and in the calculation of gaseous and liquid effluents l monitoring instrumentation alarm / trip set points and in the conduct of the environmental monitoring program.

I. Operable See Appendix A Technical Specifications.

l l Amendment No. 1 l

J. Process Control Program (PCP)

The PCP is a document which identifies the current formulas, sampling methods, analyses, tests, and determinations used to control the processing and packaging of solid radioactive vastes. The PCP con-trols these activities in such a way as to assure compliance with 10 CFR 20, 10 CFR 61, 10 CFR 71 and other applicable regulatory require-ments governing the disposal of the radioactive waste.

K. Rated Thermal Power See Rated Power, Appendix A Technical Specifications.

L. Site Boundary The Site Boundary is that line beyond which the land is not owned, leased, or otherwise controlled by NYPA and NMPC. Refer to Figure 5.1-1 for the map of the site boundary with regard to liquid and gas-eous releases.

M. Solidification Solidification is the conversion of wet wastes into a form that meets shipping and burial ground requirements.

N. Source Check A Source Check is the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

O. Treatment Any process which effectively reduces the concentration of radioac-tive material per unit measure released to the environment. This includes such processes as filtration, evaporation / condensation, set-t11ng/ decanting, and solidification.

P. Unrestricted Area An unrestricted area shall be any area at or beyond the site boundary access to which is not controlled by NYPA 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 pur-poses.

The definition of unrestricted area used -in implementing the Radio-logical Effluent Technical Specifications has been expanded over that in 10 CFR 20.3(a)(17). The unrestricted area boundary may coincide with the exclusion (fenced) area boundary, as defined in 10 CFR 100.3(a), but the unrestricted area does not include areas over water bodies. The concept of unrestricted areas, established at or beyond the site boundary, is utilized in the Limiting Conditions for Opera-tion to keep levels of radioactive materials in liquid and gaseous effluents as low as is reasonably achievable, pursuant to 10 CFR 50.30a.

Amendment No. 2

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 2.0 LIQUID EFFLUENTS 2.1 LIQUID EFFLUENT MONITORS 2.1 LIQUID EFFLUENT MONITORS Applicability Applicability Applies to instrumentation required for moni- Applies to instrumentation for monitoring ra-toring -radioactive liquid effluent discharges dioactive liquid effluent di charges.

to the environment as specified in Table 2.1-1.

Objective Objective To assure that radioactive liquid effluent dis- To ensure that instrumentation required for ra-charges are properly monitored and recorded dioactive liquid effluent discharges are main-during release, tained and calibrated.

Specifications Specifications

a. The limiting conditions for operation of the a. The alarm / trip set points of these channels instruments that monitor radioactive liquid shall be determined and adjusted in accor-effluents are given in Table 2.1-1. With a dance with the methodology and parameters in radioactive liquid effluent monitoring in- the Offsite Dose Calculation Manual (ODCM).

strumentation channel alarm / trip set point less conservative than required by the ODCM, without delay suspend the release of radio-active liquid effluents monitored by the affected channel, or declare the channel in-operable, or change the set point so it is acceptably conservative.

b. With less than the minimum number of radio- b. The surveillance requirements for the radio-active liquid effluent monitoring instrumen- active liquid effluent monitoring instrumen-tation channels operable, take the action tation is shown on Table 3.10-2.

shown in Table 2.1-1. Take corrective ac-

! tions to return the instruments to operable Amendment No. 3 i

4 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS ,

status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperabil-ity was not corrected in a timely manner.

NOTE: This reporting requirement does not apply to instruments which won 1' not have been required to be oper ble during the 30 day period.

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i Amendment No. 4

TABLE 2.1-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum Channels Instrument Operable Action Gross radioactivity monitors providing alarm and automatic termination of release Liquid radwaste effluent line 1 (a) i Gross beta or gamma radioactivity monitors providing alarm but not providing automatic termination of release Service water system effluent line 1 (b) l l

Flow rate measurement devices

! Liquid radwaste effluent line 1 (c) l NOTES FOR TABLE 2.1-1 I l I

(a) With the number of operable channels less than the required minimum num-l ber, effluent releases may continue provided that prior to initiating a l release

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a. Two independent samples are analyzed;

b. Two technically qualified members of the facility staff verify the i i

discharge line valving; Otherwise, suspend release of radioactive effluents via this pathway.

l (b) With the number operable of channels less than the required minimum num-ber, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a limit of detection of at least 5 x 10 7 microcuries/ml.

l l (c) With the number of operable channels less than the required minimum num-ber, effluent releases via this pathway may continue provided the flow rate is estimated at least once per four hours during actual releases.

l Pump curves or tank level decreases generated in situ may be used to es-timate flow.

l Amendment No. 5

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 2.2 CONCENTRATION OF LIQUID EFFLUENTS 2.2 CONCENTRATION OF LIQUID EFFLUENTS Applicability Applicability Applies to the concentration of radioactive ma- Applies to the analysis of radioactive liquid terials in liquid effluents. wastes from the plant through a liquid pathway to an unrestricted area.

Objective Objective To ensure that the concentrations of radioac- To ensure that analyses are performed and con-tive materials in liquid effluents are kept to centration determined for radioactive liquid acceptable levels, releases.

Specifications Specifications

a. The concentration of radioactive materials a. Radioactive liquid wastes shall be sampled released to the unrestricted areas shall and analyzed according to the sampling and not exceed the values specified in 10 CFR analyses program of Table 2.2-1.

20, Appendix B. Table II, Column 2. For dissolved or entrained noble gases the concentration shall be limited to 2x10 4 pCi/ml.

b. With the concentration of radioactive mate- b. The results of the radioactivity analyses rial released from the plant to urrestricted shall be used in accordar.ce with the methods areas exceeding the above linits, restore in the ODCM to assure that the concentra-the concentration to within the above lim- tions at the point of release are maintained its or terminate the release. within the limits of Specification 2.2.a.

Amendment No. 6

TABLE 2.2-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimum of Detection Liquid Release Sampling Analysis Type of Activity (LLD)(*)

Type Frequency Frequency Analysis (pCi/ml)

Batch Waste elease Prior to Principal gamma 5 x 10

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Tanks ) each Each batch batch emitters I#)

~0 I-131 1 x 10 One batch Monthly Dissolved and 1 x 10 -5 per month entrained gases (gamma emitters)

Quarterly H-3 1 x 10 -5 batch composite ( )

Gross alpha

-7 1 x 10 r rt -8 Quarterly Sr-89, Sr-90 5 x 10 batch composite Fe-55 1 x 10 -5

~7 Continuous Weekly Principal gamma 5 x 10 Releases (*) Continuous ( ) composite ( } emitters (#)

I-131 1 x 10 -6 Monthly Monthly Dissolved and 1 x 10 -5 grab sample entrained gases (gamma emitters)

H-3 1 x 10 -5 Continuous ( } Monthlycomposite('

-7 Gross alpha 1 x 10

-8 Sr-89, Sr-90 5 x 10 Continuous (f)composite Quarterly f) g

-5 Fe-55 1 x 10 Amendment No. 7

NOTES FOR TABLE 2.2-1 (a) The LLD (Lower Limit of Detection) is defined, for purpose of these specifications, 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 and with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement cystem (which may include radiochemical separation):

4.66 s b LLD =

6 E V 2.22 x 10 Y exp (-AAt)

Where:

LLD is the a priori lower limit of detection, as defined above (in micro-

curies per unit mass or volume);

s is the standard deviation of the backgrour.d counting rate or of the b

counting rate of a blant~ samples da appropriate (in counts per minute);

E is the counting efficiency (in counts per disintegration); .

V is the sample size (in units of mass or volume);

2.22 x 106 is the number of disintegrations per minute per microcurie; Y is the fractional radiochemical yield (when applicable);

A is the radioactive decay constant for the particular radionuclide; and at for plant efficents is H el.ipsed time be*waen the midpoint of sample ecliection and i:ime of co n.ir;.

Typica3 values of E, V, Y, and at 2hould be used in the calculation.

It should be recognized chat the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and

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not as an a posteriori (af ter the fact) limit for a particular measure-nent.

, (b) A batch ralease is the discharge o) liqu42 wastes of a discrete volume.

Prior to sampling for analyces, each batch shall be isolated, and then mixed to assure representative sampling.

(c) The principal gamma emitters for which the LLD specification applies ex-clusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-134, and Cs-137. This list does not mean that only these nuclides are to be detected and reported. Other peaks that are measurable and identifiable, together with the above nuclides, shall also be identified and reported in the Semiannual Radioactive Effluent Release Rep 9rt. The LLD for Mo-99, Ce-141, and Ce-144 is 5x10 5, ,

Amendment No. 8

NOTES FOR TABLE 2.2-1 (continued)

(d) /. 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.

(e) 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.

J (f) To be' representative of the quantities and concentrations of radioactive materials in liquid effluents, sarples 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.

L y .1 Amendment No. 9

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 2.3 DOSE FROM LIQUID EFFLUENTS 2.3 DOSE FROM LIQUID EFFLUENTS Applicability Applicability Applies to radiation doses from liquid efflu- Applies to the calculation of the radiation ents containing radioactive materials. dose from liquid effluents containing radioac-tive materials.

Objective Objective To assure that the dose limitations of 10 CFR To ensure that the radiation dose from radioac-50, Appendix I for liquids are met. tive liquid effluents is determined.

Specifications Specifications

a. The dose to a member of the public from ra- a. Cumulative dose contributions from liquid dioactive materials released from the plant effluents shall be determined in accordance in liquid effluents to unrestricted areas with the ODCM at least monthly for the cur-shall be-limited as follows: rent calendar quarter and the current calen-

, dar year.

1. During any calendar quarter, limited to less than or equal to 1.5 mrem to the whole body and to less than or equal to 5 mrem to any organ; and,

2. During any calendar year, limited to less

.than or equal to 3 mrem to the whole body and to less than or equal to 10 arem to any organ.

b. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, the fo'-

lowing shall be done:

Amendment No. 10

o LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

1. Identify the causes for such release rates;.

2. Define and initiate a program of correc-tive action; and

3. Prepare and submit a report to the NRC within 30 days.

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, LIMITING CONDITIONS FOR OPERATION -

SURVEILLANCE REQUIREMENTS I

2.4 -LIQUID RADI0 ACTIVE WASTE TREATHDIT SYSTEM' 2.4 LIQUID RADI0AC1TJ' WASTE l TREATMENT SYSTEM OPER-OPERATIONS ATIONS Applicability Applicability Applies to the operability of radioactive 11 9- Dose projections apply to liquid effluents re-uid processing equipment. leased to unrestricted areas.

Objective ' Objective l

To ensure liquid radwaste treatment system (s) To ensure that action levels to require opera-

-are operated to prevent exceeding the dose lim- tion of vast- . eatment systems are determined.

its of Specification 2.3.

Specifications Specifications i

a. The liquid radioactive waste treatment sys- a. Doses to individuals in unrestricted areas tem shall be used when the . projected dose due to liquid releases shall be projected at

, from untreated liquid releases, over a 31 least monthly in accordance with ODCM. .

l day period, to a member of the public would

, exceed:

1. 0.06 arem to the whole body; or, 4

2. 0.2 mrem to any organ. [

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b. With radioactive liquid waste being dis-charged in excess of the'above limits, pre-pare and submit to the Connaission within 30 days a report that includes the following .

information:

{ 1. Explanation if liquid radwaste was.being discharged without treatment; and if so: i I

-Amendment No. 12 i 4

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

2. Identification of any inoperable equip-ment or subsystems and the reason for the inoperability;

3. Action (s) taken to restore the inoperable equipment to operable status; and

4. Summary description of action (s) taken to prevent a recurrence.

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Amendment No. 13 i

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 2.5 MAXIMUM ACTIVITY IN OUTSIDE TANKS 2.5 MAXIMUM ACTIVITY IN OUTSIDE TANKS Applicability Applicability Applies to tanks located outdoors that do not Applies to outdoor tanks that do not have catch have catch basins that drain back to the build- basins that drain back to the building.

ing.

Objective Objective To ensure that in the event of an uncontrolled To ensure that the radioactivity contained in release of the tank's contents, the resulting outdoor tanks is kept within applicable limits.

concentrations would be less than the limits of 10 CFR 20 Appendix B Table II, Column 2, at the nearest surface water supply in an un-restricted area.

Specifications Specifications

a. The quantity of liquid radioactive material a. The quantity of radioactive material con-

contained in a condensate storage tank or tained in a condensate storage tank or any any outside temporary tank shall be limited outside temporary tank shall be determined to 10 curies, excluding Tritium and dis- by analyzing a liquid sample of the tank's solved or entrained noble gases. contents weekly when radioactive liquid is l being added to the tank.

b. With the quantity of liquid radioactive ma-terial in a tank above this limit, reduce the tank's radioactive contents to within the limic within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ; and

c. Describe the events leading te this condi-tion in the next Semiannual Effluent Release Report.

Amendment No. 14

BASES 2.0 LIQUID EFFLUENTS 2.1 LIQUID EFFLUENT MONITORS The radioactive liquid effluent instrumentation is provided to monitor and control the releases of radioactive materials in liquid effluents during planned or unplanned releases. The alarm / trip set points for these instruments shall be calculated in accordance with methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the lim-its of 10 CFR 20. The operability and use of this instrumentation is consistent with the requirements of 10 CFR 50, Appendix A General Design Criteria 60, 63 and 64.

2.2 CONCENTRATION OF LIQUID EFFLUENTS This specification is provided to ensure that the concentration of radio-active materials released in liquid waste effluents to unrestricted areas will be less than the concentration levels specified in 10 CFR 20, Appen-dix B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposure above (1) the design objectives of 10 CFR 50, Appendix I, Section II.A. to a member of the public and (2) the limits of 10 CFR 20 Appendix B. Table II, Column 2 to the population.

The concantration limit for dissolved or entrained noble gases is based on Xe-135 as the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in ICRP Publication 2.

2.3 DOSE FROM LIQUID EFFLUENTS This specification is provided to assure that the requirements of 10 CFR 50, Appendix I, Section II.A, III.A and IV.A are met. Tha Limiting Con-ditions for Operation assures that the guides set forth in Appendix I,Section II.A are met. The specifications provide the required operating flexibility and, at the same time, implement the guides set forth in Appendix I,Section IV.A, to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable."

2.4 LIQUID RADI0 ACTIVE WASTE TREATMENT SYSTEM OPERATIONS 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 specification assures that the requirements of 10 CFR 50.36a, 10 CFR 50, Appendix A. General Design Critcrion 60, and design objective of 10 CFR 50, Appendix I, Section II.D are met. 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 10 CFR 50, Appendix I, Section II.A for liquid effluents.

Amendment No. 15

BASES 2.5 MAXIMUM ACTIVITY IN OUTSIDE TANKS Restricting the quantity of radioactive material contained in the speci-fied tanks provides assurance that, in the event of an uncontrolled re-lease of the tank's contents, the resulting soncentrations would be less than the limits of 10 CFR 20, Appendix B, Table II, Column 2, at the nearest drinking water supply currently in use and at the nearest surface p water supply in an unrestricted area.

Amendment No. 16

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.0 GASEOUS EFFLUENTS 3.1 GASEOUS EFFLUENT MONITORS 3.1 GASEOUS EFFLUENT MONITORS Applicability Applicability These requirements apply to instruments which Applies to instrumentation listed in Specifi-monitor the gaseous radioactivity effluent cation 3.1.a and analyses of gaseous effluent pathways to the environment, releases.

Objective Objective To assure that radioactive gaseous effluent To ensure that instrumentation required for discharges are properly monitored and recorded gaseous effluent releases is maintained and during release, calibrated and the radioactivity of gaseous re-leases is determined.

Specifications Specifications

a. Radioactive gaseous wastes released to the a. Operation of the gaseous effluent monitors environment via the below listed pathways listed in Specification 3.1.a shall be veri-shall be monitored and recorded during re- fled by performing instrument surveillance lease from the respective pathway. as specified on Table 3.10-2.

1. Main stack

2. Rcfuel floor vent

3. Reactor building vent

4. Turbine building vent

5. Radwaste building vent Amendment No. 17 l

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

b. Each pathway listed in a. above, shall also b. The iodine cartridge and the particulate be sampled for iodine and particulate radio- filter for each pathway listed in Specifi-activity on a continuous basis during re- cation 3.1.a shall be changed out at least lease from the respective pathway. weekly.

c. If Specifications 3.1.a and b. , above, can- c. Grab samples, when required, shall be col-not be met, effluent releases may continue lected at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and ana-via the respective pathway provided gaseous lyzed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of collection. Auxil-grab samples are collected in the case of a lary samplers shall run continuously and be monitor out of service and auxiliary sam- changed out at least weekly, plers are used in case a particulate or io-dine sampler is out of service;

1. Return the instrument to operable status within 30 days; or

2. Provide an explanation in the next Semi-annual Radioactive Release Report as to why the inoperability was not corrected within 30 days.

d. Alarm / trip set points shall be determined in accordance with the ODCM and set to ensure that the limits of Specification 3.2 are not exceeded. With a radioactive gaseous ef-fluent monitoring instrumentation channel alarm / trip set point less conservative than required by the above specification:

1. Without deld;, 3uspend the release of radioactive gaseous effluents monitored l by the affected channel; or

2. Declare the channel inoperable; or

3. Change the set point so it is acceptably conservative.

1 Amendment No. 18

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.2 CASEOUS DOSE RATES 3.2 CASEOUS DOSE RATES Applicability Applicability Applies to the radiation dose from radioactive Applies to the calculation of the dose rates raterial in gaseous effluents from the plant. from radioactive materials in gaseous effluents from the plant.

Objective Objective To ensure that the dose rates at or beyond the To ensure that appropriate calculations are site boundary from gaseous effluents do not performed to. determine the dose rates from exceed the annual dose limits of 10 CFR 20 for gaseous effluents from the plant.

unrestricted areas.

Specifications Specifications

a. The dose rate at or beyond the site boundary a. The dose rate due to noble gases in gaseous due to radioactive materials released from effluents shall be determined to be contin-the plant in gaseous effluents shall be lim- uously within the limits of Specification ited as follows: 3.2.a. in accordance with the methods and procedures of the ODCM.

1. 5500 mrem / year to the whole body and

. 53000 mrem / year to the skin from noble gases; and,

2. 51500 mrem / year to any organ from Iodine-131, Iodine-133. Tritium and for radioac-

]' tive materials in particulate form with half-lives greater than 8 days (inhala-tion pathway only),

b. With the dose rate (s) exceeding the above b. The dose rate due to Iodine-131. I:di==-133, limits, without delay restore the release Tritium and to radionuclides in particulate rate to within the above limits. form with half-lives greater than 8 days in Amendment No. 19

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS gaseous effluents, shall be determined to be

'within the above limits in accordance with the methods and procedures of the ODCM. This will be done by obtaining representative samples and performing analyses in accor-dance with the sampling and analyses program specified in Table 3.2-1.

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i Amendment No. 20

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i TABLE 3.2-1 RADIOACTIVE CASE 0US WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimum of Detection Gaseous Release Sampling Analysis Type of Activity (LLD)(a)

Type Frequency Frequency Analysis (pCi/ml)

Main Stack and Monthly Monthly Principal Gamma 1 x 10 '

Refuel Floor Vent Grab Noble Emitters ( }

and Sample (d) g,,,,(b)

Reactor Building Vent and Quarterly Quarterly H-3 1 x 10 Turbine Building Grab Sample Vent and

-2 Radwaste Building Continuous ( Weekly I-131 1 x 10 Vent Charcoal Sample * '

-1 Continuous (# Weekly Principal Gamma 1 x 10 Particulate Emitters Sample * * (I-131, others)

-11 Continuous (*) 1 Wk/Mo Gross Alpha 1 x 10 Particulate Sample

-11 Continuous (*) 4 Wk/Qr Sr-89, Sr-90 1 x 10 Composite Particulate Sample Continuous ("} Noble Gas Noble Gases 1 x 10 -5 Monitor Gross Beta or Gamma Amendment No. 21

NOTES FOR TABLE 3.2-1 (a) The LLD is defined..for p'trpose of these specifications, 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% probabil-ity and with only 5% probability of falsely concluding that a blank ob-servation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation):

4.66 s b LLD =

E V 2.22 x 106 , Y + exp (-AAt)

Where:

LLD is the a priori Iower limit of detection, as defined above (in micro-curies per unit mass or volume);

s is the standard deviation of the background counting rate or of the b

counting rate of a blank sample, as appropriate (in counts per minute);

E is the counting efficiency (in counts per disintegration);

V is the sample size (in units of mass or volume);

2.22 x 100 is the number of disintegrations per minute per microcurie; Y is the fractional radiochemical yield (when applicable);

A is the radioactive decay constant for the particular radionuclide; and at for plant effluents is the elapsed time between the midpoint of sample collection and time of counting.

Typical values of E, V, Y, and at should be used in the calculation.

It should be recognized that the LLD 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.

i (b) The principal gamma emitters for which the LLD specification applies ex-clusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, Xe-135m, and Xe-138 for gaseous emissions; and, Mn-54, Fe-59, l Co-58, co-60, 2n-65, Mo-99, Cs-134, and Cs-137 for particulate emissions.

This list does not mean that only these nuclides are to be detected and reported. Other peaks that are measurable and identifiable, together

with the above nuclides, shall also be identified and reported in the Semiannual Radioactive Effluent Release Report. The LLD for Mo-99, Ce-141, and Ce-144 is 5x10 II.

(c) The ratio of the sample flow rate to the sampled stream flow rate shall te known for the time period covered by each dose or dose rate salcula-tion made in accordance with Specifications. This determinatioc shall be made using design flow rates if flow meters are not provided or are inoperable.

Amendment No. 22

NOTES FOR TABLE 3.2-1 (continued)

(d) Main stack sempling and analysis shall also be performed following shut-down, start-up, or a thermal power change exceeding 20% of reted thermal power in one hour. This requirement applies only if: (1) analysis shows that the dose equivalent I-131 concentration in the primary coolant has increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent activity has increased by more than a factor of 3 af ter correcting for increases due to changes in thermal power level in l both cases.

1 (e) When the offgas filter system charcoal absorbers are not in service, main stack samples shall be changed at least once per 7 days and analyses

! shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months af ter changing, or af ter removal from sampler. In addition, sampling shall be performed following each shut-down, start-up or thermal power change exceeding 20% of rated thermal power in one hour. In those instances, sampling shall be performed at least daily until two consecutive samples show no increase in concentra-tian but in no case for more than 7 consecutive days. This requirement applies only if: (1) analysis shows that the dose equivalent I-131 con-centration in the primary coolant has increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has increased more than a factor of 3 after correcting for increases due to changes in thermal power level in both cases. Analysis of daily sample; shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 l

hours are analyzed, the corresponding LLDs may be increased by a factor of 10.

(f) When the offgas filter system charcoal absorbers are in service, amin stack samples shall be changed at least once per 7 days and analyses shall be completed within 48 hears after changing, or after removal from j sampler. In addition, sampling shall be performed following each shut-down and start-up. In those instances, sampling shall be performed at least daily until two consecutive samples show no increase in concentra-tion but in no case for more than 7 consecutive days. This requirement applies only if: (1) analysis shcws that the dose equivalent I-131 con-centration in the primary coolant has increased more than a factor of 3; and (2).the noble gas monitor shows that effluent activity has increased more than a factor of 3 after correcting for increases due to' changes in thermal power level in both cases. Analysis of daily samplea shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLDs may be increased by a factor of 10.

l l

l Amendment No. 23 l

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.3 AIR DOSE, NOBLE CASES 3.3 AIR DOSE. NOBLE CASES Applicability Applicability 4 Applies to the air dose due to noble f,ases re- Applies to the calculation of the air dose due leased from the plant. to noble gas effluent.

Objective Objective To assure that the noble gas dose liaitations To ensure that appropriate calculations are of 10 CFR 50. Appendix I, are met, performed to determine the air dose from noble gas effluents.

Specifications Specifications r

a. The air dose to areas at or beyond the site a. Cumulative air dose contributions for noble boundary from noble gases released from the gases shall be calculated at least monthly plant in gascom effluents shall be limited: in accordance with the ODCM for the current calendar quarter and the current calendar

1. During any c lendar quarter, to less than year.

or equal to 5 mrad from gamma radiation, and less than or equal to 10 mrad from beta radiation; and,

, 2. During any calendar year, to less than

or equal to 10 mrad from gamma radiation and less than or equal to 20 mrad from beta radiation.

b. With the calculated air dose from radioac-tive noble gases in gaseous effluents ex-

, ceeding any of the above limits, prepare and

- submit to the Commission, within 30 days, a report that:

Amendment No. 24

-LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

1. Identifies the cause(s) for exceeding the limit (s); and

2. Defines the corrective actions that have been taken to reduce the releases; and

3. Identifies the proposed corrective ac-tions to be taken to assure than subse-quent releases will be in compliance with the above limits.

Amendment No. 25

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.4 DOSE DUE TO IODINE-131, 10 DINE-133. TRITIUM AND 3.4 DOSE DUE TO IODINE-131. IODINE-133. TRITIUM AND RADIONUCLIDES IN PARTICULATE FORM = RADIONUCLIDES IN PARTICUIATE FORM Applicability Applicability Applies to the cumulative dose from Iodine-131, Applies to the calculation of the dose due to 2 Iodine-133, Tritium, and radionuclides in par- Iodine-131 Iodine-133 Tritium, and radionu-ticulate form in gaseous effluents. clides in particulate form in gaseous efflu-ents.

Objective Objective l To assure that the dose limitations of 10 CFR To ensure that appropriate calculations are 1

50, Appendix I, are met. performed to determine the dose from Io-dine-131 Iodine-133, Tritium, and radionu-clides in particulate form.

Specifications Specifications

a. The dose to a member of the public at or a. Cumulative dose contributions shall be cal-beyond the site boundary from Iodine-131, culated at least monthly in accordance with Iodine-133 Tritium, and radionuclides in the ODCM for the current calendar quarter particulate form with half-lives greater and the current calendar year.

than 8 days released from the plant in gaseous effluents shall be limited:

1. During any calendar quarter to less than or equal to 7.5 mrem to any organ; and,

2. During any calendar year to less than or equal to 15 mrem to any organ.

9 1 b. With the calculated dose from the release of r Iodine-131, Iodine-133. Tritium, and radio-nuclides in par :iculate form with half-lives greater than 8 days, in gaseous effluents Amendment No. 26

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS exceeding any of the above limits, prepare and submit to the Commission within 30 days a report that:

1. Identifies the cause(s)-for exceeding the limit; and

2. Defines the corrective actions that have been taken to reduce the releases; and

3. Identifies the proposed corrective ac-tions to be taken to assure that subse-quent releases will be in compliance with the above limits.

Amendment No. 27

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.5 MAIN CONDENSER STEAM JET AIR EJECTORS (SJAE) 3.5 MAIN CONDENSER STEAM .ET AIR EJECTORS (SJAE) b Applicability Applicability Applies to main condenser offgas discharge rate Applies to the point of discharge at the SJAE.

for noble gases.

Objective Objective To ensure that the SJAE release rates are main- To ensure that the SJAE release rates are prop-tained at a level compatible for further treat- erly monitored.

ment and release.

Specifications Specifications

a. The gross radioactivity (beta and/or gamma) a. The gross radioactivity (beta and/or gamma) rate of noble gases measured at the SJAE is rate of noble gases from the SJAE shall be given on Table 3.10-1. determined to be within the limits of Speci-fication 3. 5. r. by performing an isotopic analysis of a representative sample of gases taken at the discharge (prior to dilution and/or discharge) of the SJAE as follows:

1. At least monthly.

2. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following an increase as indicated by the SJAE Monitor, of greater than 50% (after factoring out increases due to changes in thermal power level) in the nominal steady state fission gas re-lease from the primary coolant.

1 Amendment No. 28

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

b. Except as specified in 1. below, both SJAE b. Operation of the SJAE radioactive offgas system radiation monitors shall be operable monitors shall be verified by performing during reactor power operation. The trip instro.ent surveillance as specified on time delay setting for closure of the SJAE Table 3.10-2.

isolation valve shall not exceed 15 min.

1. In the event that one of the two SJAE radiation monitors is made or found to be inoperable, continued reactor power operation is permissible provided that the inoperable monitor is tripped in the downscale position.

2. Upon the loss of both SJAE system radia-tion monitors, either temporarily monitor radiation levels at the SJAE or initiate an orderly shutdown and have the main steam isolation valves closed within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Amendment No. 29

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6 0FFCAS TREATMENT SYSTEM 3.6 0FFCAS TREATMENT SYSTEM Applicability Applicability Applies to the system installed for reduction Applies to the calculation of the radiar. ion of radioactive materials in gaseous waste prior dose from gaseous effluents containing radioac-to discharge. tive materials.

Objective Objective To minimize concentration of radioactive mate- To ensure that treatment of gaseous wastes by rials released from the site. the offgas system is implemented when required.

Specifications Specifications

a. The offgas treatment system shall be used a. With the offgas treatment system not in use, to reduce the concentration of radioactive doses due to gaseous releases from the site materials in gaseous effluents prior to shall be projected at least monthly in ac-release from the plant within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after cordance with the ODCM.

the start-up of the second turbine driven feedwater pump.

b. With gaseous radwaste from the main condens-er air ejector system being discharged with-out treatment for more than 7 days when the system is required to be operable, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that includes the following informa-tion:

1. Identification of the causes, including any inoperable equipment or subsystems, and the reason for their inoperability; Amendment No. 30

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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 0FFCAS TREATMENT SYSTEM EXPLOSIVE CAS MIXTURE 3.7 0FFGAS TREATMENT SYSTEM EXPLOSIVE CAS MIXTURE INSTRUMENTATION INSTRUMENTATION Applicability Applicability Applies to the condenser offgas treatment sys- Applies to the offgas treatment system instru-tem recombiner operation. mentation, which monitors the critical oper-ating parameters of the primary recombiner.

4 Objective Objective To ensure proper conditions for the offgas re- To ensure that instrumentation required for au-combiner to operate at design efficiency in tomatic isolation is maintained and calibrated, order to prevent an explosive mixture of gases in the charcoal treatment system.

Specifications Specifications

a. The concentration of either hydrogen or oxy- a. The concentration of either hydrogen or oxy-gen in the main condenser of gas treatment gen in the main condenser offgas treatment system shall be limited to less than or system shall be determined to be within the equal to 4% by volume. limits of Specification 3.7 a by continuous-ly monitoring the waste gases in the main condenser offgas treatment system whenever the main condenser evacuation system is in operation with the hydrogen or oxygen moni-tors. Operation of the hydrogen or oxygen monitors shall be verified in accordance with Specification 3.7.b.1, 2 and 3.

b. In lieu of continuous hydrogen or oxygen b. Whenever continuous hydrogen or oxygen mon-monitoring, the following $nstrumentation itoring is not available, operation of the shall be operationai and cep61e of pro- explosive gas mixture instruments listed in viding automatic isolation o' the offgas Specification 3.7.b shall be verified.

Amendment No. 32

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS treatment system under the following condi- 1. An instrument check shall be performed tions: daily when the offgas treatment system is in operation.

1. The offgas dilution steam flow instrumen-tation shall alarm and automatically iso- 2. An instrument channel functional test late the primary offgas recombiner at low shall be performed once per operating flow less than 6000 pounds per hour or cycle, high flow greater than 7200 pounds per hour. 3. An instrument channel calibration shall be performed once per operating cycle.

2. The offgas inlet temperature sensor shall alarm and automatically isolate the pri-mary offgas recombiner a temperature less than 125*C.

3. The primary offgas recombiner outlet tem-perature shall alarm and automatically isolate the offgas treatment system on a temperature less than 150*C.

c. In lieu of continuous hydrogen or oxygen c. With condenser offgas treatment system re-monitoring, the condenser offgas treatment combiner in service, the hydrogen content system recombiner shall be analyzed to veri- shall be verified weekly to be 54% by vol-fy that it contains less than or equal to 4% ume.

hydrogen by volume.

In the event that the hydrogen content can-

d. With the requirements of the above spec- not be verified, operation of this system ifications not satisfied, restore the may continue for up to 14 days.

recombiner system to within operating speci-fications or suspend use of the charcoal treatment system within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Amendment No. 33

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.8 STANDBY CAS TREATMENT SYSTEM (SBGTS) 3.8 STANDBY CAS TREATMENT SYSTEM (SBGTS)

Applicability Applicability Applies to the SBGTS instrumentation. Applies to the surveillance requirement of the instrumentation which activates the SBGTS.

Objective Objective To assure that the SBCTS is actuated with the To specify the instrument surveillance type and proper signal, frequency.

Specifications Specifications The limiting conditions for operation are given The instrument surveillance requirements are on Table 3.10-1.

given on Table 3.10-2.

Amendment No. 34

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.9 MECHANICAL VACUUM PUMP ISOLATION 3.9 MECHANICAL VACUUM PUMP ISOLATION Applicability Applicability Applies to the mechanical pump isolation in- Applies to the surveillance requirement which strumentation. isolates the mechanical vacuum pump.

Objective Objective To assure operability of the trip circuitry. To specify the instrument surveillance type and frequency.

Specifications Specifications a..The mechanical vacuum pump shall be capable The instrument surveillance requirements are of being automatically isolated and secured given on Table 3.10-2.

by a signal of high radiation in the main steam line tunnel whenever the main steam isolation valves are open.

b. If the limits of Table 3.10-1 are not met, the vacuum pump shall be isolated.

Amendment No. 35

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.10 MAIN CONTROL ROOM VENTILATION RADIATION MONITOR 3.10 MAIN CONTROL ROOM VENTILATION RADIATION MONITOR Applicability Applicability Applies to the emergency air supply treatment Applies to the instrumentation which monitors system to the main control room. the control room supply air.

Objective Objective To assure operability of the in-line radiation To specify the instrument surveillance type and monitor and annunciator. frequency.

Specifications Specifications The limiting conditions for operation are given The instrument surveillance requirements arc on Table 3.10-1. given on Table 3.10-2.

Amendment No. 36

TABLE 3.10-1 RADIATION MONITORING SYSTEMS THAT INITIATE AND/OR ISOLATE SYSTEMS Minimum No. Total Number of of Operable Instrument Channels Provided by Design Instrumegg) Trip Level Setting for Both Channels Action Channels Trip Function 1 Refuel Area Exhaust Monitor (b) 2 (c) or (d) 1 Reactor Building Area Exhaust (b) 2 (d)

Monitors 1 SJAE Radiation Monitors $500,000 pCi/sec 2 (e) 1 Turbine Building Exhaust Monitors (b) 2 (f) 1 Radwaste Building Exhaust Monitors (b) 2 (f) 1 Main Control Room Ventilation 54 x 10 8 cpm (I 1 (g)

Monitor 2 Mechanical Vacuum Pump Isolation 53 x Normal Full 4 (h) rever Background NOTES FOR TABLE 3.10-1 (a) Whenever the systems are required to be operable, there shall be two operable or tripped instrument channels per trip system. From and after the time it is found that this cannot be met, the indicated action shall be taken.

(b) Trip level setting is in accordance with the methods and procedures of the ODCM.

(c) Cease operation of the refueling equipment.

(d) Isolate secondary containment and start the SBGTS.

(e) Bring the SJAE release rate within the limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in hot standby within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

(f) Refer to Appendix B LCO 3.1.d.

(g) Control room isolation is manually initiated.

i (h) Uses same sensors as primary containment isolation on high main steam line radiation. Refer to Appendix A Table 3.2-1.

(i) Conversion factor is 8.15 x 107 cpm - 1 pCi/cc.

Amendment No. 37

TABLE 3.10-2 MINIMUM TEST AND CALIBRATION FREQUENCY FOR RADIATION MONITORING SYSTEMS (*}

Instrument Instrument Channel Instrument Channel Logic System Func-Instrument Channels Check (b) Functional Test (I) Calibration (I) tion Test (f)(h)

Main Stack Exhaust Monitors Daily Quarterly Quarterly --

Refuel Area Exhaust Monitors Daily Quarterly Quarterly --

Reactor Building Area Exhaust Moni- Daily Quarterly Quarterly Semiannually tors / Isolation Turbine Building Exhaust Monitors Daily Quarterly Semiannually --

Radwaste Building Exhaust Monitors Daily Quarterly Semiannually --

SJAE Radiation Monitors /Offgas Line Daily Quarterly Quarterly Semiannually Isolation Main Control Room Ventilation Monitor Daily Quarterly Quarterly --

Mechanical Vacuum Pump Isolation (E} -- - --

Once Per Oper-ating Cycle Liquid Radwaste Discharge Monitor / Iso-Daily When Quarterly Quarterly Semiannually lation ( )(d)(e)(f) Discharging LiquidRadwasteDisgrgeFlowRate Daily Quarterly Once Per Oper- --

Measuring Devices ating Cycle LiquidRadgteDischargeRadioactivity Daily Quarterly Once Per Oper- -

Recorder ating Cycle Normal Service Water Effluent Daily Quarterly Quarterly -

SBGTS Actuation -- -- --

Semiannually Amendment No. 3S

NOTES FOR TABLE 3.10-2 (a) Functional tests, calibrations and instrument checks need not be per-formed when these instruments are not required to be operable or are tripped.

(b) Instrument checks shall be performed at least once per day during these periods when the instruments are required to be operable.

(c) A source check shall be performed prior to each release.

(d) Liquid radwaste effluent line instrumentation surveillance requirements need not be performed when the instruments are not required as the result of the discharge path not being utilized.

(e) An instrument channel calibration shall be performed with known radioac-tive sources standardized on plant equipment which has been calibrated with NBS traceable standards.

(f) Simulated automatic actuation shall be performed once each operating cycle. Where possible, all logic system functional tests will be per-formed using the test jacks.

(g) Refer to Appendix A tor instrument channel functional test and instrument channel calibration requirements (Tables 4.1-1 and 4.1-2 respectively).

These requirements are performed as part of main steam high radiation monitor surveillances.

l (h) The logic system functional tests shall include a calibration of time delay relays and timers necessary for proper functioning of the trip systems.

(1) This instrumentation is excepted from the functional test definition.

The functional test will consist of inj ecting a simulated electrical signal into the measurement channel. These instrument channels will be calibrated using simulated electrical signals once every three months.

l l

l Amendment No. 39

i BASES 3.0 GASEOUS EFFLUENTS 3.1 GASEOUS EFFLUENT MONITORS The radioactive gaseous effluent instrumentation is provided to monitor and control the releases of radioactive materials in gaseous effluents during planned or unplanned releases. The alarm / trip set points for these instruments shall be calculated in accordance with methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the lim-its of 10 CFR 20.

The operability and use of this instrumentation is consistent with the requirements of 10 CFR 50, Appendix A, General Design Criteria 60, 63 and 64.

3.2 GASEOUS DOSE RATES This specification is provided to ensure that the dose at or beyond the site boundary from gaseous effluents will be within the annual dose lim-its of 10 CFR 20. The annual dose limits are the doses associated with the concentrations of 10 CFR 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharges in gaseous effluents will not result in the exposure of a member of the public to annual average concentrations exceeding the limits specified in 10 CFR 20, Appendix B Table II (10 CFR 20.106[b]). The specified limits restrict, at all times, corresponding gamma and beta dose rates above background to an individual at or beyond the exclusion area boundary to

$500 mrem / year to the total body or to 53000 mrem / year to the skin. These limits also restrict the corresponding thyroid dose rate above background to a child via the inhalation pathway to $1500 mrem / year.

3.3 AIR DOSE, NOBLE GASES This specification is provided to assure that the requirements of 10 CFR 50, Appendix I, Section II.B. III.A and IV.A are met. The Limiting Con-ditions for Operation are the guiden set forth in Appendix I,Section II.B. The specification provides the required operating flexibility and, at the same time, implements the guides set forth in Appendix I,Section IV.A. to assure that the releases of radioactive material in gaseous ef-fluents will be kept "as low as is reasonably achievable."

3.4 DOSE DUE TO 10 DINE-131, IODINE-133, TRITIUM AND RADIONUCLIDES IN PARTIC-ULATE FORM This specification is provided to assure that the requirements of 10 CFR 50, Appendix I, Section II.C, III.A and IV.A are met. The Limiting Con-ditions for Operation are the guides set forth in Appendix I, Section II.C. The specifications provide the required operating flexibility and, at the same time, implement the guides set forth in Appendix I, Section Amendment No. 40

BASES IV.A. to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable."

3.5 MAIN CONDENSER STEAM JET AIR EJECTOR (SJAE)

This specification is provided to assure that remedial action is taken to limit the noble gas release rate at the SJAE. The requirement provides reasonable assurance that the amount of noble gas that must be treated and/or released is controlled to a level that prevents exceeding the lim-its specified in 10 CFR 20, Appendix B. Table II.

Two air ejector offgas monitors are provided and when their trip point is reached, cause an isolation of the air ejector offgas line. Isolation is initiated when both instruments reach their high trip point or one has an upscale trip and the other a downscale trip. There is a 15 minute delay before the air ejector offgas isolation valve is closed. This delay is accounted for by the 30 minute holdup time of the offgas before it is released to the stack. Both instruments are required for trip but the instruments are so designed that any instrument failure gives a downscale trip.

3.6 0FFGAS TREATMENT SYSTEM This specification is provided to ensure that the syste:n will be avail-able for use when required to reduce projected doses due to gaseous releases. This specification assures that the requirenents of 10 CFR 50.36a, 10 CFR 50 Appendix A, General Design Criterion 60, and design objective in 10 CFR 50 Appendix I, Section II.D are met. The specified limits governing the use of appropriate portions of the systems are spec-ified as a suitable fraction of the guide values set forth in 10 CFR 50, Appendix I, Sections II.B and II.C. for gaseous effluents.

The requirement for offgas treatment system operability provides assur-ance that the releaea of radioactive materials in gaseous waste will be kept "as low as is reasonably achievable." Operability of the system is based upon start-up of the second turbine driven feedwater pump. This is due to the fact that excess air in-leakages in the main condenser as a result of operating only one turbine driven feedwater pump will exceed offgas treatment system limitations and consequently render the system inoperabic. Start-up of the second turbine driven feedwater pump will decrease air in-leakages and assure offgas treatment system availability.

3.7 0FFCAS TREATMENT SYSTEM EXPLOSIVE CAS MIXTURE INSTRUMENTATION This specification is provided to ensure that the concentration of poten-tially explosive gas mixtures contained in portions of the offgas treat-ment system not designed to withstand a hydrogen explosion is maintained below the lower explosive limit of hydrogen. The proper operation of the primary recombiner ensures that the charcoal contained in the condenser Amendment No. 41

l I

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BASES l offgas treatment system is not exposed to an explosive mixture of gases.

Thus it provides assurance that the releases of radioactive materials 1 will be controlled in conformance with the requirements of 10 CFR 50, Appendix A, General Design Criterion 60.

3.8 STANDBY GAS TREATMENT SYSTEM (SBGTS)

Four radiation monitors are provided which initiate isolation of the reactor building and operating of the SBGTS. The monitors are located as follows: two in the reactor building ventilation exhaust duct and two in refuel floor ventilation exhaust duct. Each pair is considered a separats system. The trip logic consists of any upscale trip on a single monitor or a downscale trip on both monitors in a pair to cause the desired ac-tion.

Trip settings for the monitors in the refueling area ventilation exhaust ducts are based upon initiating normal ventilation isolation and SBGTS operation so that most of the activity released during the refueling ac-cident is processed by the SBGTS.

The radiatien monitors in the refueling area ventilation duct which ini-tiate building isolation and standby gas treatment operation are arranged in a one out of two logic system. The bases given in Appendix A Bases 4.2 for the rod blocks apply here also and were used to arrive at the func-tior..1 testing frequency. The air ejector offgas monitors are connected in a two out of two logic arrangement. Based on experience with instru-unts of similar design, a testing interval of once every three months has been found adequate.

3.9 MECHANICAL VACUUM PUMP ISOLATION l

3.10 MAIN CONTROL ROOM VENTILATION RADIATION MONITOR Amendment No. 42

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 4.0 SOLID RADIOACTIVE WASTE 4.1 PROCESS CONTROL PROGRAM 4.1 PROCESS CONTROL PROGRAM Applicability Applicability Applies to radioactive solid waste packages for Applies to the solidification system utilized i offsite chipment. for wet solid wastes. .

Objective Objective i To ensure that solid radioactive vaste meet 10 To ensure that solidification of wet solid i CFR 20, 10 CFR 61 and 10 CFR 71 shipping and wastes is performed in accordance with the PCP.

burial ground requirements prior to shipping.

Specifications specifications

a. The solid radwaste system shall be used in a. The PCP shall be used to verify the solidi-accordance with the PCP to process wet ra- fication of at least one representative test dioactive wastes to meet shipping and burial specimen from at least every tenth batch of i

ground requirements. each type of wet radioactive waste.

j 1. If any test specimen fails to verify so-1 lidification, the solidification of the batch under test shall be suspended until the following are completed:

4 l a) Additional test specimens can be ob-tained; b) Alternative solidification parameters can be determined in accordance with the PCP; and 1

Amendment No. 43 i

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS c) A subsequent test verifies solidifi-cation.

Solidification of the batch may then be resumed using the parameters determined by the PCP.

2. If the initial test specimen from a batch of waste fails to verify solidification, the PCP shall provide for the collection and testing of representative test speci-mens from each consecutive batch of the same type of wet waste until at least 3 consecutive initial test specimens demon-strate solidification. The PCP shall be modified as required, to assure solidifi-cation of subsequent batches of waste,

b. With the provisions of the PCP not satis-fied, suspend shipments of affected packaged solid radioactive wastes from the site.

Amendment No. 44

BASES 4.0 SOLID RADIOACTIVE WASTE This specification assures that the requirements of 10 CFR 50.36a and 10 CFR 50, Appendix A, General Design Criterion 60 are met. The process parameters included in establishing a PCP may include, but are not lim-ited to: waste type, waste pH, waste / liquid / solidification / agent /ca-talyst ratios, waste oil content, waste principal chemical constituents, mixing and curing times.

l Amend at No. 45

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 5.0 TOTAL DOSE 5.1 TOTil. DOSE FROM URANIUM FUEL CYCLE 5.1 TOTAL DOSE FROM URANIUM FUEL CYCLE g licability Applicability Applies to radiation dose from releases of ra- Applies to the calculation of total dose due to dioactivity and radiation from uranium fuel releases of radioactivity and radiation from cycle sources, uranium fuel cycle sources.

Objective Objective To assure that the requirements of 40 CFR 190 To ensure that appropriate calculations are are met. performed to determine total dose to a member of the public.

Specifications Specifications

a. The dose or dose commitment to any member of a. Dose Calculations Cumulative dose contribu-the public, due to releases of radioactivity tions from liquid and gaseous effluents and radiation, from uranium fuel cycle shall be determined in accordance with Spec-sources shall be limited as follows: ifications 2.3.a. 3.3.a. and 3.4.a and in accordance with the ODCM.

1. Less than or equal to 25 mres/ year to the whole body; and,

2. Less than or equal to 25 mres/ year to any organ except the thyroid which shall be limited to less than or equal to 75 ares / year.

b. With the calculated doses from the release b. Cumulative dose contributions from direct of radioactive materials in liquid or gas- radiation from the reactor units and from eous effluents exceeding twice the limits radwaste storage tanks shall be determined Amendment No. 46

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS of Specification 2.3.a or 3.3.a or 3.4.a. in accordance with the methodology and pa-calculations shall be made including an rameters in the ODCM. This requirement is estimate of direct radiation contributions applicable only under conditions set forth to determine whether the limits of 5.1 have in Specification 5.1.f been exceeded. If this is the case, a report defining corrective actions to be taken to reduce subsequent releases to levels within limits, along with a schedule for achieving conformance, shall be prepared and submitted to the Commission within 30 days. This re-port, as defined in 10 CFR 20.405c, shall include estimates of the radiation exposure (dose) to a member of the public from urani-um fuel cycle sources, including all efflu-ent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report. It shall also de-scribe levels of radiation and concentration of radioactive material involved, and the cause of the exposure Icvels or concentra-tions. If the estimated dose (s) exceed (s) the above limits, and if the release condi-tion resulting in violation of 40 CFR 190 has not already been corrected, the report shall include a request for variance in ac-cordance with the. provisions of 40 CFR 190.

Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is com-plete.

Amendment No. 47

l FIGURE 5.1-1 l

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h ! (pnvate LANIVIEW NIACARA MONAWK NEW YORK POWER CORPORAfl0N POWER AUTHORITY h

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NOTES TO FIGURE 5.1-1 (a) NMP1 stack (height is 350 feet)

(b) NMP2 stack (height is 430 feet)

(c) JAFNPP stack (height is 385 feet)

(d) Building vents (ground level)*

(e) NMP1 radioactive liquid discharge (Lake Ontario, bottom)

(f) NMP2 radioactive liquid discharge (Lake Ontario, bottom)

(g) JAFNPP radioactive liquid discharge (Lake Ontario, bottom)

(h) Site boundary (1) Lake Ontario shoreline

No credit taken for the elevations of these release points and therefore treated as ground level releases.

Additional Information:

-- NMP2 reactor building vent is located 187 feet above ground level

-- JAFNPP reactor and turbine building vents are located 173 feet above ground level

-- JAFNPP radwaste building vent is 112 feet above ground level Amendment No. 49

1 I

l BASES i I

5.0 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR I 190. This specification requires the preparation and submittal of a re- l port whenever the calculated dose from plant radioactive effluents exceed !

twice the design objective doses of 10 CFR 50, Appendix I. The 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 190 lim-its. For the purpose of the report, it may be assumed the dose commit-ment to the member of the public from other uranium fuel cycle sources is negligible. However, dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be con-sidered. If the dose to any member of the public is estimated to exceed the requirements of 40 CFR 190, the report, with a request for variance (provided the release conditions resulting in a violation of 40 CFR 190 have not already been corrected), shall be submitted in accordance with provisions of 40 CFR 190.11 and 10 CFR 20.405c. This request is con-sidered a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR 190 and does not apply in any way to the requirements for dose limitation addressed in Specifications 2.0 and 3.0. An individ-ual 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 nu-clear fuel cycle.

Amendment No. 50

T LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 6.0 RADIOLOGICAL ENVIRONMENTAL MONITORING 6.1 MONITORING PROGRAM 6.1 MONITORING PROGRAM Applicability The radiological environmental monitoring sam-pies shall be collected, pursuant to Table At all times. 6.1-1, from the locations given in the table and figure (s) in the ODCM and shall be analyzed Objective pursuant to the requirements of Table 6.1-1, and the detection capabilities required by To evaluate the effects of plant operation on Table 6.1-3.

the environs and to verify the effectiveness of the controls on radioactive material.

Specifications

a. With the radiological environmental moni-toring program not being conducted as speci-fled in Table 6.1-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report, a descrip-tion of the reasons for not conducting the program as required and the plans for pre-venting a recurrence.

(Deviations are permitted from the required sampling schedule if samples are unobtain-able due to hazardous conditions, seasonal unavailability, theft, uncooperative resi-dents, or to malfunction of automatic sam-pling equipment. If the latter, efforts shall be made to complete corrective action prior to the end of the next sampling period.)

Amendment No. 51

LIMITING CCNDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

b. With the level of radioactivity (as the re-sult of plant effluents) in an environmental sampling medium at a specified location ex-ceeding the reporting levels of Table 6.1-2 when averaged over any calendar quarter, prepare and submit to the Commission a re-port within thirty (30) days from the end of the affected calendar quarter or within thirty (30) days from the time it is deter-mined that a reporting level has been ex-ceeded. This report shall identify the cause(s) for exceeding the limit (s) and define the corrective action (s) to be taken to reduce radioactive effluents so that the calculated annual dose to a member of the public is Jess than the calendar year limits of Specifications 2.3, 3.3, and 3.4. When more than one of the radionuclides in Table 6.1-2 are detected in the sampling medium, l this report shall be submitted if:

concentration (1) # concentration (2) # ****

limit level (1) limit level (2)

When radionuclides other than those in Ta-ble 6.1-2 are detected and are the result of plant effluents, this report shall be sub-mitted if the calculated annual dose to an individual is equal to or greater than the calendar year limits of Specification 2.3, 3.3, and 3.4.

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 de-scribed in the Annual Radiological Environ-mental Operating Report.

Amendment No. 52

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 6.1-1, locations for obtaining replacement samples shall be identified and added to the radiological environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program. The cause of the unavailability of samples and the new location (s) for obtaining replacement sam-ples shall be identified in the next Semi-annual Radioactive Effluent Release Report.

Also included in the report shall be a revised figure (s) and table for the ODCM reflecting the new location (s).

Amendment No. 53

TABLE 6.1-1 OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Sampling and Pathway Collectioy,) Type and Frequency and/or Sample Number of Samples (,) and Locations Frequency of Analysis AIRBORNE Radiciodine Samples from 5 locations: Continuous sam- Radioiodine Canisters:

and ple operation Analyze weekly for I-131.

Particulates a. 3 samples from offsite locations in dif- with sample col-ferent sectors of the highest calculated lection weekly Particulate Samples:

site average D/Q (based on all licensed or as required GrossbetaradioactivitTb) site reactors). by dust loading, following filter change ,

whichever is composite (bylocaylyn)

b. I sample from the vicinity of a community more frequent. for gamma isotopic having the highest calculated site aver- quarterly (as a minimum),

age D/Q (based on all licensed site re-actors).

c. I sample from a control location 9 to 20 milesdistantygjintheleastprevalent wind direction Direct 32 stations with two or more dosimeters Quarterly Gamma dose monthly or Radiation (,) placed as follows: An inner rin'g of stations quarterly.

in the general area of the site boundary and an outer ring in the 4 to 5 mile range from the site with a station in each of the land based sectors of ecch ring. There are 16 land based sectors in the inner ring and 8 land based sectors in the outer ring. The balance of the stations (8) are placed in special interest areas such as population centers, nearby residences, schools, and in 2 or 3 areas to serve as control stations.

Amendment No. 54

TABLE 6.1-1 (continued)

Exposure Sampling and Pathway Collectioy,) Type and Frequency and/or Sample Number of Samples (,) and Locations Frequency of Analysis WATERBORNE Surface ('} a. I sample upstream. Composite sam- Camma isotopic analysis ple over one monthly. Composite for

b. Isamplefromthesiyg most downstream month period 8) . Triti nalysis quar-cooling water intake terly Sediment from I sample from a downstream area with existing Twice per year. Cammaisotop{g) analysis Shoreline or potential recreational value. semiannually INGESTION Milk a. Samples from milch animals in 3 locations Twice per month. Camma isotopic and I-131 within 3.5 miles distant having the high- April through analysis twice per month est calculated site average D/Q. If December (sam- when milch animals are there are none, then I sample from milch ples will be on pasture (April through animals in each of 3 areas 3.5 to 5.0 collected in December); monthly (Jan-miles distant having the highest calcu- January through uarythryuyhMarch),if lated site average D/Q gysed on all March if I-131 required licensed site reactors) is detected in November and

b. I sample from milch animals at a control December of the location (9 to 20 miles distan preceding year).

lessprevalentwinddirection)[d9ndina Amendment No. 55

TABLE 6.1-1 (continued)

Exposure Sampling and Pathway Collectioy,) Type and Frequency and/or Sample Number of Samples ) and Locations Frequency of Analysis Fish a. I sample of each of 2 commercially or Twice per year. Gamma isotopic ("} analysis recreationally important species in the of edible portions.

vicinity of a site discharge point.

b. I sample of each of 2 species (same as in a. above or of a species with similar feedinghabits)fromanareid7tleast5 miles distant from the site Food Products a. Samples of 3 different kinds of broad Monthly when Gamma isotopic (#} analysis.

leaf vegetation (edible or inedible) available (Isotopic to include grown nearest each of two different off- (May thrcugh I-131.)

site locations of highest calculated October),

annual average ground level D/Q if milk sampling is not performed (based on all licensed site reactors).

b. I sample of each of the similar broad Monthly when Gamma isotopic (C analysis.

leaf vegetation grown 9-20 miles distant available (Isotopic to include in the least prevalent wind direction if (May through I-131.)

milk sampling is not performed. October).

c. In lieu of the garden census as specified Once, during Gamma isotopic ( } analysis, in 6.2, samples of at least 3 different harvest season. (Isotopic to include kinds of broad leaf vegetation (edible or I-131.)

inedible) may be performed at the site boundary in each of 2 different direction sectors with the highest calculated D/Qs.

I sample each of 3 similar broad leaf varieties of vegetation grown 9-20 miles distant in the I prevalent wind direction sector Amendment No. 56

NOTES FOR TABLE 6.1-1 (a) It is recognized that, at times, it may not be possible or practical 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. Actual locations (distance and directions) from the site shall be provided in the Annual Radiological Environmental Operating Report. Calculated site averaged D/Q values and meteorological parameters are based on historical data (specified in the ODCM) for all licensed site reactors.

(b) Particulate sample filters should be analyzed for gross beta 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air is greater than 10 times a historical yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

(c) Gamma isotopic analysis means the identification and quantification of gamma emitting radionuclides that may be attributable to the effluents from the plant.

(d) The purpose of these samples is to obtain background information. If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites which provide valid background data may be substituted.

(e) 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 addi-tion to, integrating dosimeters. For ti.e purpose of this table, a thermo-luminescent dosimeter may be consdered to be, one phosphor and two or more phosphors in a pocket may be considered as two or more dosimeters. Film badges shall not be used for measurin;, direct radiation.

(f) The " upstream sample" shall be taken at a distance beyond significant in-fluence of the discharge. The "dow ntream sample" shall be taken in an area beyond, but near, the mixing zone. if practical.

(g) Composite samples should be collected with equipment (or equivalent) which is capable of collecting an aliquoit at time intervals which are very short (e.g., hourly) relative to the compositing period (e.g., monthly) in order to assure that a representative sample is obtained.

(h) A milk sampling location, as required in Table 6.1-1 is defined as a location having at least 10 milking cows present at a designated milk sample location. It has been found from past experience, and as a result of conferring with local farmers, that a minimum of 10 milking cows is necessary to guarantee an adequate supply of milk twice per month for analytical purposes. Locations with less than 10 milking cows are usually utilized for breeding purposes which eliminates a stable supply of milk for samples as a result of suckling calves and periods when the adult animals are dry. In the event that 3 milk sample locations cannot meet the requirement for 10 milking cows, then a sample location having less than 10 milking cows can be used if an adequate supply of milk can reasonably and reliably be obtained based on communications with the farmer.

Amendment No. 57

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NOTES FOR TABLE 6.1-3 (a) The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability and with 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical sep-aration),

4.66 s b "E*V 2.22 + Y exp (- Aat)

Where:

LLD is the a priori lower limit of detection, as defined above (in pico-curie per unit mass or volume);

s is the standard deviation of the background counting rate or of the b

counting rate of a blank sample, as appropriate (in counts per minute);

E is the counting efficiency (in counts per transformation);

V is the sample size (in units of mass or volume);

2.22 is the number of transformation,s per minute per picoeurie; Y is the fractional radiochemical yield (when applicable);

A is the radioactive decay constant for the particular radionuclide; at is the elapsed time between sample collection (or end of the sample collection period) and time of counting.

Typical values of E, V, Y, and at should be used in the calculations.

(b) It should be recognized that the LLD 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 measure-ment. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluc-tuations, 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 identi-fied and described in the Annual Radiological Environmental Operating Report.

(c) LLD for drinking water samples. If no drinking water pathway exists, the LLD of the gamma isotopic an-11ysis may be used.

Amendment No. 60

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 6.2 LAND USE CENSUS PROGRAM 6.2 LAND USE CENSUS PROGRAM Applicability The land use census shall be conducted during the growing season at least once per 12 months At all times. using the information that will provide the best results, such as by a door to door survey, Objective aerial survey, or by consulting local agricul-ture authorities, etc. The results of the land To identify locations of milch animals and gar- use census shall be included in the Annual Ra-dens of greater than 50 square meters within 3 diological Environmental Operating Report.

miles of the site.

Specifications

a. A land use census shall be conducted and shall identify the locations of all milch animals, the nearest residence, and all gar-dens

of greater than 50 square meters pro-ducing fresh leafy vegetables, in each of the 16 meteorological sectors within a dis-tance of 5 miles from the site.

b. With a land use census identifying a milch animal in a location (s) which represents a calculated D/Q value greater than the values currently being used in calculating Surveil-lance Requirement 3.4, identify the new lo-cation (s) in the next Semiannual Radioactive Effluent Release Report.

Broad leaf vegetation sampling may be performed in lieu of the garden census as specified in Table 6.1-1.

Amendment No. 61

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS

c. With the land use census identifying a milch animal location (s) that represents a calcu-lated D/Q (via the same exposure pathway) 50% greater than at a location from which samples are currently being obtained in accordance with Table 6.1-1, add the new location (s) to the radiological environmen-tal monitoring program within 30 days. The sampling location (s), excluding the control station location, having the lowest calcu-lated D/Q (via the same exposure pathway) may be deleted from this monitoring program after October 31 of the year in which this land use census is conducted. Identify the new location (s) in the next Semiannual Ra-dioactive Effluent Release Report and in-clude the additions in the ODCM.

Amendment No. 62

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 6.3 INTERLABORATORY COMPARISON PROGRAM 6.3 INTERLABORATORY COMPARISON PROGRAM Applicability A summary of the results obtained as part of the above required Interlaboratory Comparison At all times. Program shall be included in the Annual Radio-logical Environmental Operating Report.

Objective To provide quality control of environmental sample analyses.

Specifications

a. Analyses shall be performed on radioactive materials supplied as part of an Interlab-oratory Comparison Program which has been approved by the Commission. Participation in this program shall include all media for which samples are routinely collected and for which intercomparison samples are avail-able.

b. With analyses not being performed as re-quired in 6.3.a above, report the corrective actions taken to prevent a recurrence in the Annual Radiological Environmental Operating Report.

Amendment No. 63

BASES 6.0 RADIOLOGICAL ENVIRONMENTAL MONITORING 6.1 MONITORING PROGRAM The radiological environmental monitoring program required by this speci-fication provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures to members of the public resulting from station operation. This monitoring program assures that 10 CFR 50, Appendix I, Section IV.B.2 is met. It thereby supplements the radiologi-cal effluent monitoring program by verifying that the measurable concen-trations of radioactive materials and levels of radiation are not higher than expected, based on the effluent measurements and the modeling of the environmental exposure pathways. The initial specified monitoring program will be effective for at least the first three years of commercial opera-tion. 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 Limit of Detection (LLDs). The LLDs re-quired by Table 6.1-3 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system. The LLD is not an a posteriori (after the fact) limit for a particular measurement.

6.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring program are made if required by the results of this cen-sus. The best survey information, such as that from door to door sur-veys, aerial surveys, consultations with local agricultural authorities, etc., shall be used. This census satisfies the requirements of 10 CFR 50, Appendix I, Section IV.B.3. Restricting the census to gardens of greater than 50m 2 provides assurance that significant exposure pathways via leafy vegetables will be identified 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, Revision 1, October 1977, for consumption by a child. To determine this minimum garden size, the following assumptions were made: (1) 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage),

and, (2) a vegetable yield of 2 kg/m 2. In lieu of the garden census the significance of the garden exposure pathway can be evaluated by the samp-ling of green leafy vegetables as specified in Table 6.1-1.

Amendment No. 64 ,

i 1

l BASES 6.3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an Interlaboratory Comparison Pro-gram is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in the environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

Amendment No. 65 w -

1 7.0 ADMINISTRATIVE CONTROLS 1

7.1 RESPONSIBILITY .

a. The Resident Manager shall have direct responsibility for assuring the operation of the James A. FitzPatrick Plant is conducted in such a manner as to provide continuing protection to the environment.

During periods when the Resident Manager is unavailable, he may del-egate his responsibilities to the Superintendent of Power, or in his absence, to other qualified supervisory personnel,

b. Implementation of the Radiological Effluen* Technical Specifications is the responsibility of the Superintendent of Power, with the as-sistance of the plant staff organization.

7.2 PROCEDURES Written procedures and administrative policies shall be established, im-plemented and maintained that meet or exceed the requirements and recom-mendations of Section 5 " Facility Administrative Policies and Procedures" of ANSI 18.7-1972 and Regulatory Guide 1.33, November 1972, Appendix A.

In addition, procedures shall be established, implemented and maintained for the PCP, ODCM, and Quality Control Program for effluent and environ-mental monitoring using the guidance in Regulatory Guide 4.1, Revision 1.

7.3 REPORTING REQUIREMENTS

a. Planned Liquid and Gaseous Releases ,

The limits for radioactive materials contained in liquid and gaseous effluents are contained in Specifications 2.3, 3.3 and 3.4.

b. Environmental Samples Exceeding Limits of Table 6.1-2 When the limits of Table 6.1-2 are eiceeded, refer to Specification j 6.1.b for reporting requirements.

l

c. Semiannual Radioactive Effluent Release Report l

Routine Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be sub-mitted within 60 days af ter January 1 and July 1 of each year. The i period of the first report shall begin with the date of initial criticality.

1. The Radioactive Effluent Release Report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit using as guidance Regu-latory Guide 1.21, Revision 1, June 1974, " Measuring, Evalu-ating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from

Light-Water-Cooled Nuclear Power Plants", with data summarized on a quarterly basis following the format of Appendix B there-of.

Amendment No. 66 l

l

2. The Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year may include an annual sum-mary of meteorological data collected over the previous year.

If the meteorological data is not included, the licensee shall retain it on file and provide it to the U.S. Nuclear Regulatory Commission upon request. This same report shall include an as-sessment of the radiation doses

due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year to the public. All assumptions.used in making these assessments (i.e., specific activity, expceure time and location) shall be included in these reports. The assessment of radiation doses shall be performed in accordance with the ODCM.

3. The Radioactive Effluent Release Reports shall include any change to the PCP or the ODCM made during the reporting period, as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 6.2.

4. The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year 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) during the previous calendar year, to show conformance with 40 CFR 190, Environmental Radiation Pro-tection Standards for Nuclear Power Operation. This assessment of radiation doses is performed in accordance with the ODCM.

I

5. The Radioactive Effluent Release Reports shall include the following information for each class of solid waste (defined by 10 CFR 61) shipped offsite during the report period:

! (a) Container volume; l (b) Total curie quantity (specify whether determined by mea-surement or estimate),

i l (c) Principal radionuclides (specify whether determined by measurement or estimate),

(d) Source of waste and processing employed (e.g., dewatered l spent resin, compacted dry waste, evaporator bottoms),

(e) Type of container (e.g., LSA, Type A, Large Quantity), and (f) Solidification agent or absorbent (e.g., cement, Dow media, etc.).

l

6. The Radioactive Effluent Release Reports shall include a list and description of unplanned releases, to unrestricted areas of radioactive materials in gaseous and liquid effluents made dur-ing the reporting period.

!

The dose assessment sections of the Semiannual Radiological Effluent Release

! Report shall be submitted within 90 days af ter January 1 of each year as an l addendum to the Semiannual Radiological Effluent Release Report.

Amendment No. 67

7. The Radioactive Effluent Release Report shall contain the cause for unavailability of any environmental sample required by Table 6.1-1 and shall identify the locations for obtaining re-placement samples. This shall also include a revised figure (s) and table for the ODCM reflecting the new location (s). Refer to Specification 6.1.c.

8. The Radioactive Effluent Release Report shall contain new loca-tions identified in the land use census in accordance with Specifications 6.2.b or 6.2.c.

9. The Radioactive Effluent Release Report shall contain the events leading to the condition which resulted in exceeding 10 curies for tanks specified in the Limiting Conditions for Operation, Section 2.5.a

d. Annual Radiological Environmental Operating Report Routine Radiological Environmental Reports covering the operation of the unit during the pervious calendar year shall be submitted prior to May 1 of each year.

The Annual Radiological Environmental Operating Reports shall in-clude summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period. The report shall include a comparison with preoperational studies, operational controls (as appropriate), and environmental surveillance reports from the previous five years, 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 Specification 6.2.

The Annual Radiological Environmental Operating Reports shall in-clude the results of analysis of all radiological environmental samples and of all measurements taken during the period pursuant to Table 6.1-1, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radio-logical Assessment Branch Technical Position, Revision 1. November 1979. In the event that some individual results are not available for inclusion in the report, the report shall note and explain 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 maps

covering all sampling locations and keyed to a table giving distances and directions from the centerline of the reactor; the results of participation in the Interlaboratory Comparison Pro-gram required by Specification 6.3 (or appropriate EPA cross-check program code), and discussion of all analyses in which the LLD's re-quired by Table 6.1-3 were not routinely achievable.

One map shall cover stations near the site boundary; a second shall include the more distant stations.

Amendment No. 68

ATTACHMENT II to JPN 86 PROPOSED CHANGES TO APPENDIX A ACCOMPANYING THE RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS (RETS)

(JPTS-83-09)

NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59

JAFNPP TABLE OF CONTENTS (cont'd)

Page D. Emergency Service Water System 240 E. Intake Deicing Heaters 242 3.12 Fire Protection Systems 244a 5.0 Design Features 245 5.1 Site 245

5.2 Reactor 245 5.3 Reactor Pressure Vessel 245 5.4 Containment 245 5.5 Fuel Storage 245 5.6 Seismic Design 246 6.0 Administrative Controls 247 6.1 Responsibility 247 6.2 Plant Staff Organization 247 6.3 Plant Staff Qualifications 248 6.4 Retraining and Replacement Training 248 6.5 Review and Audit 248

6. 5.1 ' ' Plant Operating Review Committee (PORC) 248a 6.5.2 Safety Review Committee (SRC) 250 6.6 Reportable Occurrence Action 253 6.7 Safety Limit Violation 253 l 6.8 Procedures 253 6.9 Reporting Requirements 254a 6.10 Record Retention 254g

6.11 Radiation Protection Program 255 l

6.12 Industrial Security Program 258 6.13 Emergency Plan 258 6.14 Fire Protection Program 258 6.15 Environmental Qualification 258a l

l i Amendment No. X, X. g , )(,JT iii e

i

i-JAFNPP TABLE OF CONTENTS (cont'd)

Page 6.16 Process Control Program (PCP) 258b i 6.17 Offsite Dose Calculation Manual (ODCM) 258b 6.18 Major Modifications to Radioactive 258c Liquid, Gaseous, and Solid Waste Treatment Systems 7.0 References 285 Amendment No. )',)MI IV

JAFNPP LIST OF TABLES Table Title Page 3.1-1 Reactor Protection System (Scram) 41 Instrumentation Requirement 4.1-1 Reactor Protection System (Scram) 44 Instrument Functional Tests 4.1-2 Reactor Protection System (Scram) 46 Instrument Calibration 3.2-1 Instrumentation that Initiates Primary 64 Containment Isolation 3.2-2 Instrumentation that Initiates or controls 66 the Core and Containment Cooling Systems 3.2-2 Instrumentation t. hat Initiates control 72 Rod Blocks 3.2-4 (DELETED) 74 l 3.2-5 Instrumentation that Monitors Leakage 75 Detection Inside the Drywell 3.2-6 Surveillance Instrumentation 76 3.2-7 Instrumentation that Initiates 77 Recirculation Pump Trip 4.2-1 Minimum Test and Calibration Frequency 78 for PCIS 4.2-2 Minimum Test and Calibration Frequency 79 for Core and Containment Cooling System 4.2-3 Minimum Test and Calibration Frequency 81 for Control Rod Blocks Actuation 4.2-4 (DELETED) 82 l

4.2-5 Minimum Test and Calibration Frequency 83 for Drywell Leak Detection 4.2-6 Minimum Test and Calibration Frequency 84 for Surveillance Instrumentation i 4.2-7 Minimum Test and Calibration Frequency 86 1 for Recirculation Pump Trip l l

l Amendment No. )WP v j l

i .

i JAFNPP l

}

! Z. Top of Active Fuel The Top of Active Fuel, corresponding

! to the top of the enriched fuel

! , column of each fuel bundle, is i located 352.5 inches above vessel zero, which is the lowest point in

, the inside bottom of the reactor vessel. (See General Electric drawing No. 919D690BD.)

l i

AA. Rod Density 1

! Rod density is the number of control I rod notches inserted expressed as a fraction of the total number of control rod notches. All rods fully 4 inserted is a condition representing i

100 percent rod density.

j I -

AB. Puree-Purcina Purge or Purging is the controlled l process of discharging air or gas from l a confinement in such a manner that i replacement air or gas is required to l

purify the confinement.

AC. Ventine Venting is the controlled process of releasing i air or gas from a confinement in such a manner that replacement air or gas is not provided or required.

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Amendment No. 75 6a l

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JAFNPP 3.2 (cont'd) 4.2 (cont'd) controls are required to be operable System logic shall be functionally as specified in Specification 3.5. tested as indicated in Table 4.2-2.

C. Control Rod Block Actuation C. Control Rod Block Actuation

1. The limiting conditions of Instrumentation shall be functionally operation for the instrumen- tested, calibrated, and checked as tation that initiates control as indicated in Table 4.2-3.

rod block are given in Table 3.2-3. System logic shall be functionally tested as indicated in Table 4.2-3.

2. The minimum number of operable instrument channels specified in Table 3.2-3 for the rod block monitor may be reduced by one in one of the trip systems for main-tenance and/or testing, provided that this condition does not last longer than 24 hrs. in any 30 day period.

D. Radiation Monitoring Systems - D. Radiation Monitoring Systems -

Isolation & Initiation Functions Isolation and Initiation Functions Refer to the Radiological Effluent Refer to the Radiological Effluent Technical Specifications Technical Specifications (Appendix B). (Appendix B).

6 Amendment No. 50

Pages 51-53 INTENTIONALLY DELETED I

Amendment No. 51

JAFNPP 3.2 Bases (cont'd) the specifiction are adequate to assure the above criteria are met. The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration, and also minimizes the risk of inadvertent operation; i.e., only one instrument chaanel out of service.

Flow integrators are used to record the integrated flow of liquid from the. dry-well sumps. The alarm unit in each integrator is set to annunciate before the valves specified in Specification 3.6.D are exceeded.

For each parameter monitored, as listed in Table 3.2-6, by comparing the reading of each channel to the reading on redundant or related instrument channel a near continuous surveillance of instrument performance is available. Any deviation in readings will initiate any early recalibration thereby maintaining the quality of the instrument readings.

Amendment No. JHf 59

JAFNPP 4.2 Bases (cont'd) channel. Bypassing both channels for 2. More than one channel should simultaneous testing should be avoided, not be bypassed for testing at any one time.

The most likely case would be to stipulate that one channel be bypassed, tested, and restored, and then immediately following, the second channel be bypassed, tested, and restored. This is shown by Curve No. 4.

Note that there is no true minimum. The curve does have a definite knee and very little reduction in system unavailability is achieved by testing at a shorter interval than computed by the equation for a single channel.

The best test procedure of all those examined is to perfectly stagger the tests. That is, if the test interval is The automatic pressure relief four months, test one or the other instrumentation can be considered to be channel every two months. This is shown a 1 out of 2 logic system and the bases '

in Curve No. 5. The difference between given above for the rod blocks apply cases 4 and 5 is negligible. There may here also and were used to arrive at be other arguments, however, that more the functional testing frequency.

strongly support the perfectly staggered tests, including reduction in human error.

The conclusions to be drawn are these:

1. A 1 out of n system may be treated the same as a single channel in terms of choosing a test interval; and Amendment No.)( 63

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! JAFNPP 3.6 and 4.6 BASES (cont'd)

\

! The samples include Charpy V-notch iodine activity would not be j specimens of base metal, and heat expected to change rapidly over a i affected zone metal. These samples period of 96 hr. In addition, the I will receive neutron exposure more trend of the stack offgas release

! rapidly than the vessel wall rate, which is continuously material and, therefore, will lead monitored, is a good indicator of the wall in integrated neutron flux the trend of the iodine activity in exposure. The measurements of the reactor coolant. Also during .

neutron flux and testing of selected reactor startups and large power groups of specimens at intervals changes which could affect iodine j over the lifetime of the reactor may levels, samples of reactor coolant l be used to confirm the predicted shall be analyzed to insure iodine j affects on NDTT. concentrations are below allowable I levels. Analysis is required

! C. Coolant Chemistry whenever the I-131 concentration is

within a factor of 100 of its j A radioactivity concentration limit allowable equilibrium value. The

of 20 pCi/al total iodine can be necessity for continued sampling reached if the gaseous effluents are following power and offgas near the limit as set forth in transients will be reviewed within i Radiological Effluent Technical 2 years of initial plant startup.

! Specification Section 3.2a if there l is a failure or a prolonged shutdown The surveillance requirements i of the cleanup domineralizer. 4.6.C.1 may be satisfied by a i j continuous monitoring system capable 1 In the event of a steam line rupture of determining the total iodine '

i outside the drywell, with this coolant concentration in the coolant on a activity level, the resultant real time basis, and annunciating at radiological dose at the site bound- appropriate concentration levels i ary would be 33 ren to the thyroid, such that sampling for isotopic under adverse meteorological condi- analysis can be initiated. The tions assuming no more than design details of such a system must ;

3.1 pCi/gm of dose equivalent I-131. be submitted for evaluation and l The reactor water sample will be accepted by the Commission prior to itsl

used to assure that the limit of implementation and incorporation in i j Specification 3.6.C is not these Technical Specifications.

1 exceeded. The total radioactive r

)

i Amendment No. 148 t

i JhFNPP 3.7 (cont'd) 4.7 (cont'd) i i

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3. The containment shall be purged 3. Continuous Leak Rate Monitor

! through the Standby Gas Treatment

! System whenever the primary con-

! tainment integrity is required. If this requirement cannot be met, then purging shall be discontinued without delay.

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Amendment No. JP 176

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8. Review the Emergency-Plan and implementing procedures annually.

4

9. Perform special review and/or investigations at the request of the Resident Manager.

, 10. Review of those reportable occurrences requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification to the NRC, in accordance with Specification 6.9. !

, 11. Review the Offsite Dose Calculation Manual (ODCM) and implementing procedures at least once per 24 months.

i

12. Review the Process Control Program (PCP) at least once per 24 months.

4 (F) Authority 1

The PORC shall function to advise the Resident Manager on all

matters related to nuclear safety and environmental operations.

l The PORC shall recommend approval or disapproval to the Resident Manager of those items considered in 6.5 lE (1) through (4) and ;

determine if items considered in 6.5 lE (1) through (5) constitute unreviewed safety questions, as defined in 10 CFR 50.59, 4

j Ia.the event of a disagreement between the PORC and the Resident Manager, the Chairman of the SRC and the Executive Vice President-Nuclear Generation or their designated alternatives, shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and written notification provided on the next business day; however, the Resident Manager shall have responsibility for resolution of such disagreement pursuant to section 6.1.

(G) Records Minutes of all meetings of the PORC shall be recorded and 4

numbered. Copies will be retained in file. Copies will be

forwarded to the chairman of the SRC and the Executive Vice i President-Nuclear Generation.

(H) Procedures conduct of the PORC and the mechanism for implementation of its responsibilities and authority are defined in the pertinent Administrative Procedures.

6.5.2 SAFETY REVIEW COMMITTEE (SRC) t FUNCTION 6.5.2.1 The SRC shall collectively have the competence required to review problems in the following areas:

a. Nuclear power plant operations

b. Nuclear engineering

c. Chemistry and radiochemistry

+ d. Metallurgy i e. Instrumentation and control Amendment No. pd,,fs 250

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- . . . , ., , . . . . _ _ , , _ _ - _ , . _ . . . _ _ , , _ _ _ _ . . _ , . ~ , . . . . _ , _ , _ . , _ . , . . . _ , . _ _ , . . . . . . , , - . , _ . . - , . _ _ , _ , . . , . - . .,_m.,

c. The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation that

, affect nuclear safety at least once per 6 months,

d. The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix "B", 10 CFR 50, at least once per 24 months.

e. The Facility Emergency Plan and implementing procedures at lease once per 12 months.

4

f. The Facility Security Plan (including the safeguards contingency Plan) and implementing ,

procedures at least once per 12 months.

1 g. Any other area of facility operation considered appropriate by the SRC or the Executive Vice President-Nuclear Generation.

i

h. The Facility Fire Protection Program and

~

implementing procedures at least once per two years,

i. An independent fire protection and loss of prevention inspection and audit shall be performed annually utilizing either qualified offsite licensee personnel an outside fire protection. firm.

4

j. An inspection and audit of the fire protection and loss prevention program shall be performed l by an outside qualified fire consultant at i intervals no greater than 3 years.

i k The Radiological Environmental Monitoring Program and the results thereof at least once per 12 months.

1. The Offsite Dose Calculation Manual and-implementing procedures at least once per 24 months, l

m. The Process Control Program and implementing procedures for processing and packaging of radioactive wastes at least once per 24 months.

l- n. The performance of activities required by the Quality Assurance Program to. meet the provisions of Regulatory Guide 1.21, Revision 1, June 1974 and Regulatory Guide 4.1, Revision 1, April 1975 at least.once per 12 months.

6.5.2.9 AUTHORITY 4 The SRC shall report to and advise the Executive Vice President-Nuclear Generation on those areas-of responsibility

! specified in Section 6.5.2.7 and 6.5.2.8.

Amendment No. JAf, fdf, fr'I, Jti 252a

6.5.2.10 RECORDS Records will be maintained in accordance with ANSI 18.7-1972. The following shall be prepared, approved and distributed as indicated l below:

a. Minutes of each SRC meeting shall be prepared, approved and forwarded to the Executive Vice President-Nuclear Generation within 14 days after the date of the meeting.

b. Reports of review encompassed by Section 6.5.2.7 above shall be prepared, approved and forwarded to the Executive Vice President-Nuclear Generation within 14 days following completion of the review.

c. Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to the Executive Vice President-Nuclear l Generation and to the management positions responsible for the areas audited within 30 days after completion of the audit.

Amendment No. 252b l

I CHARTER l 6.5.2.11 Conduct of the committee will be in accordance with a charter approved by the Executive Vice President-Nuclear Generation setting forth the mechanism for implementation of the committee's !

responsibilities and authority.

6.6 REPORTABLE OCCURRENCE ACTION (A) In the event of a Reportable Occurrence, the NRC shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.

(B) Each Reportable Occurrence requiring 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification to the NRC shall be reviewed timely by the PORC and a report submitted by the Resident Manager to the Executive Vice President-Nuclear Generation and the SRC.

4 6.7 SAFETY LIMIT VIOLATION (A) If a safety limit is exceeded, the reactor shall be shut down and reactor operation shall only be resumed in accordance with the provisions of 10 CFR 50.36 (c)(i).

! (B) An immediate report of each safety limit violation shall be made to the NRC by the Resident Manager. The

Executive Vice President-Nuclear Generation and Chairman

! of the SRC will be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(C) The PORC shall prepare a complete investigation report of each safety limit violation and include appropriate analysis and evaluation of: (1) applicable circumstances preceding the occurrence; (2) effects of the occurrence upon facility component systems or structures; and (3) corrective action required to prevent recurrence. The Resident Manager shall forward this report to the Executive Vice President-Nuclear Generation, Chairman of the SRC and the NRC.

l l 6.8 PROCEDURES (A) Nritten procedures and administrative policies shall be established, implemented and maintained that meet or exceed the requirements and recommendations of Section 5 " Facility Administrative Policies and Procedures" of ANSI 18.7-1972 and i Appendix A of Regulatory Guide 1.33, November 1972. In addition, procedures shall be established, implemented and maintained for the Fire Protection Program and other programs, as specified in Appendix B of the Radiological Effluent Technical Specifications, Section 7.2.

(B) Those procedures affecting nuclear safety shall-be reviewed by_PORC and approved by the Resident Manager prior to implementation.

(C) Temporary changes to nuclear-related procedures may be made provided:

1. The intent of the original procedure is not altered.

Amendment No. ,56, 54r, k8J, Js 253

(A) ROUTINE REPORTS AND REPORTABLE OCCURRENCES (Continued)

4. REPORTABLE OCCURRENCES (Continued) 4.1 g. Conditions arising from natural or man-made events that, as a direct result of the event require plant shutdown, operation of safety systems, or other protective measures required by technical specifications,

h. Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the technical specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.

i. Performance of structure, system, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the safety analysis report or technical specifications bases; or discovery during plant life of conditions not specifically considered in the safety analysis report or technical specifications that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.

j. Occurrence of radioactive material contained in liquid holdup tanks in excess of that permitted by the Limiting conditions for Operation established in the Technical Specifications.

I k. Offeite releases of radioactive material i

in liquid and gaseous effluents which exceed the limits of Sections 2.3, 3.3, or 3.4 in Appendix B of the Radiological Effluent Technical Specifications.

1 Amendment No.JWr 254-e I

6.16 PROCESS CONTROL PROGRAM (PCP)

A. The PCP shall be a manual containing operational information concerning the solidification of radioactive wastes from liquid systems.

B. The PCP shall be maintained at the plant consistent with these Technical Specifications and with approved plant procedures.

C. Revisions of the PCP:

1. shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the revisions were made effective. This submittal shall contain:

a. sufficiently detailed information to support the rationale for the revisions without benefit of additional information;

b. a determination that the revision did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and

c. documentation that the revision has been reviewed and found acceptable by the PORC.

2. shall become effective upon issue following review and acceptance by the PORC.

6.17 OFFSITE DOSE CALCULATION MANUAL (ODCM)

A. The ODCM shall describe the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the l

calculation of gaseous and effluents monitoring

ins'rumentation alarm / trip setpoints consistent with l the applicable LCOs contained in these Technical

! Specifications.

B. The ODCM shall be maintained at the plant and shall l

reflect accepted methodologies and calculational procedures, l

Amendment No. 258b l

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C. Revisions of the ODCM:

1. shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the revisions were made effective. This submittal shall contain:

a. sufficiently detailed information to support the rationale for the revisions without benefit of additional information (information submitted shall consist of revised pages of the ODCM, with each page numbered and previded with an approval and date box, together with appropriate evaluations justifying the revisions);

b. a determination that the revisions will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and

c. documentation that the revisions have been reviewed and found acceptable by the PORC.

2. shall become effective upon issue following review and acceptance by the PORC.

6.18 MAJOR MODIFICATIONS TO RADIOACTIVE LIQUID, GASEOUS AND SOLID WASTE TREATMENT SYSTEMS

A. Major modifications to radioactive waste systems (liquid, gaseous and solid):

1. shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the modification is completed and made operational.

The discussion of each modification shall contain:

a. a summary of the evaluation that led to the determination that the modification could be made in accordance with 10 CFR 50.59;

b. sufficient information to support the reason for the modification without benefit of additional or supplemental information; and

c. a description of the equipment, components and processes involved and the interfaces with other plant systems.

The Authority may select to submit the information called for in this Specification as part of the annual FSAR Update.

Amendment No. 258c

2. The following evaluations shall be reported to the~

Commission in the Semiannual Radioactive Effluent Release Report, where such evaluations are required to be performed in order to assure compliance with the requirements of 10 CFR 50.59:

a. an evaluation of the modification, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto.

b. an evaluation of the modification, which shows expected maximum exposures to individuals in the UNRESTRICTED AREA and to the general population that differ from those previously estimatsd in the license application and amendments thereto; and

c. a comparison of the predicted release of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the modifications are to be made.

Amendment No. 258d

A A

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ATTACHMENT III TO JPN-84-86 SAFETY EVALUATION FOR PROPOSED RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS (RETS)

AND ASSOCIATED CHANGES (JPTS-83-09) i f

i k

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i NEW YORK POWER AUTHORITY

, JAMES A. FITZPATRICK NUCLEAR POWER PLANT l DOCKET NO. 50-333 DPR-59

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i~ I. Description of the Chances Appendix B Chances The proposed RETS include or expand upon existing radiological specifications in Appendix B. Subsection 2.3 of Appendix B,

" Radioactive Discharge", is replaced by RETS Sections 2.0 (Liquid

, Effluents) and 3.0 (Gaseous Effluents). Subsection 4.3 of Appendix B, " Radiological Environmental Monitoring," is replaced by RETS Section 6.0 (Radiological Environmental Monitoring). Section 5.0 of Appendix B, " Administrative Controls," is replaced by RETS Section 7.0 (Administrative Controls). Non-radiological specifications in Appendix B were previously deleted by Amendment

, No. 73 (Reference 2).

The two organization charts now in Appendix B (Figures 5.2-1 and 5.2-2) are being deleted by the proposed RETS.

The proposed RETS also contains additional sections which are not now included in Appendix B. Section 4.0 (Solid Radioactive Waste) addresses the " Process Control Program" (PCP) for the stabilization and solidification of radioactive wastes, and Section 5.0 (Total Dose) which includes total dose calculations due to releases ofA radioactivity and radiation from uranium fuel cycle sources.

site boundary map describing gaseous and liquid release point has been included in Section 5.0. Furthermore, RETS references an Offsite Dose Calculation Manual (ODCM) for calculating offsite doses. FitzPatrick's ODCM and PCP will be submitted to the Commission for approval in a separate submittal.

Appendix A Chances l

i Several specifications in Appendix A are added, deleted or changed

~

to accommodate the proposed RETS. Two definitions, " Purge-Purging" and " Venting", are added to the definitions section on page 6a.

In Subsection 3.2.D on pages 50-53, all references to the Radiation Monitoring Systems - Isolation & Initiation Functions including Tables 3.2-4 & 4.2-4 on pages 74 and 82 are deleted. Accordingly, the portions of the Bases addressing these items on pages 59 and 63 are deleted. These references are now addressed in the proposed RETS.

Section 4.6 Bases, page 148, reference the present Environmental Technical Specifications. This will be revised to reference the proposed RETS. In addition, " Commission" will replace "AEC". ,

i The limiting conditions for operation of the Containment Systems, Section 3.7.A.3 page 176, will read as follows:

3.7.A.3 The containment shall be purged through the l

! Standby Gas Treatment System whenever the i primary containment integrity is required. If

} this requirement cannot be met then purging shall be discontinued without delay."

111-1

I In subsection 6.5.1.E on page 250, requirements for review of the ODCM and the FitzPatrick PCP (as addressed in RETS Section 4.0) are j added to the responsibilities of the Plant Operating Review !

Committee (PORC). l In subsection 6.5.2.8 on page 252a and a new page 252b, the Safety Review Committee (SRC) responsibilities were revised to include audits of the following items:

- the Radiological Environmental Monitoring Program, 4 - the Offsite Dose Calculation Manual,

- the Process Control Program, and,

- the performance of activities required by the Quality Assurance Program to meet the provisions of Regulatory Guides 1.21 and 4.1 In subsection 6.5.2.10, on new page 252b, the word "whall" will be corrected to "shall". Additionally, in subsection 6.5.2.10c,

" Senior" will be corrected to " Executive" as previously revised by Amendment No. 78.

Reference is made, in Subsection 6.8A on page 253, to procedures specified in RETS Subsection 7.2. Two new reporting requirements, for radioactivity in liquid holdup tanks and for offsite releases,

[ are added on page 254e.

Three new subsections (6.16, 6.17 and 6.18), regarding the Process Control Program, changes to the Offsite Dose Calculation Manuals, and major modifications to radioactive waste treatment systems, are ,

added to the Administrative Controls section of Appendix A.

! Lastly, pages lii, iv, and v of the Table of Contents have been revised to reflect the proposd changes to Appendix A above and correct editorial changes to achieve consistency throughout the Specifications.

II. Purpose of the Chances The proposed RETS (Appendix B) and accompanying changes to Appendix A of the Technical Specifications are proposed to provide additional assurance that the requirements of 10 CFR 20 and 10 CFR 50, Appendix I, that radiological releases and doses be kept "as low as reasonably achievable", are met.

The proposed changes to Appendix A pages lii, iv, v. 148, and 252b are purely editorial changes to achieve consistency throughout the Technical Specifications.

! III. Impact of the Chances The proposed RETS and accompanying changes to Appendix A are designed to further insure the health and safety of the public by requiring that radioactive affluent releases and offsite doses be kept as low as is reasonably achievable. In many cases, RETS imposes radiological monitoring, analysis, control and reporting requirements over and above those now specified in Appendix B.

III-2

The Commission has provided guidance concerning the application of the standards for making a "no significant hazard considerations" determination by providing certain examples in the Federal Register (FR) Vol. 48, No. 67 dated April 6, 1984, page 14870. The proposed changes all match at least one of these examples.

The changes to Appendix B and accompanying changes to Appendix A are considered to match Commission example (vii), "A change to make l a license conform to regulations where the license change results j d

in very minor changes to facility operations." It is, therefore. '

determined to involve no significant hazard considerations. ;

l The proposed changes to pages iii, iv, v, 148, and 252b are considered to match Commission example (i), "A purely administrative change to the Technical Specifications; for example, a change to achieve consistency throughout the Technical Specifications, correction of an error, or a change in nomenclature."

IV. Implementation of the Chances l Implementation of the changes, as proposed, will not adversely impact the ALARA or Fire Protection programs at FitzPatrick.

Moreover, the changes will not adversely impact the environment.

V. Conclusion The incorporation of these changes: a) will not increase the probability or the consequences of an accident or malfunction of

equipment important to safety as previously evaluated in the Safety Analysis Report; b) will not increase the possibility of an '

accident or malfuntion of a type other than that evaluated previously in the Safety Analysis Report; c) will not reduce the margin of safety as defined in the bases for any Technical Specification; and d) does not constitute an unreviewed safety question; and, e) involves no significant hazards considerations as defined in 10 CFR 50.92.

i VI. References

1) PASNY Letter, J.P. Bayne to D. B. Vassallo, dated l February 10, 1983 (JPN-83-13).

2) NRC Letter, D. B. Vassallo to L. W. Sinclair, dated March 11, 1983 ([[::JAF-83-089|JAF-83-089]]).

3) Proposed Radiological Effluent and Monitoring Technical Specifications, submitted by PASNY letter P. J. Early to T. A. Ippolito, dated May 2, 1979 (JPN-79-26).

, III-3 l

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4) Revised Technical Evaluation Report, " Comparison of Plant and

.Model Radiological Effluent Technical Specifications,"

transmitted by NRC letter D. B. Vassallo to L. W. Sinclair, dated August 25, 1982 ([[::JAF-82-216|JAF-82-216]]).

5) NRC NUREG-0473, Rev. 3 Draft 7", " Standard Radiological Effluent Technical Specifications for Boiling Water Reactors,"

ie0ued January 1983.

6) PASNY letter, J.P. Bayne to D. B. Vassallo, dated November 22, 1982 (JPN-82-85).

7) PASNY letter, J. P. Bayne to D. B. Vassallo, dated February 10, 1983 (JPN-83-12).

8) PASNY letter, J. P. Bayne to D. B. Vassallo, dated March 9, 1983 (JPN-83-20).

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ATTACHMENT IV TO JPN-84-86

.- AD'DITIONAL INFORMATION CONCERNING THE DIFFERENCES BETWEEN THE REVISED RETS AND NUREG-J473 Revision 3 DRAFT 7" (JPTS-83-09)

C NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59

Introduction On August 8, 1984, a meeting between the NRC staff and Authority personnel was held to discuss the previous RETS submittal, dated April 29, 1983 (JPN-83-37), and comparing them to the latest revision of the Standard Radiological Effluent Technical Specification (SRETS), NUREG-0473 Rev. 3 Draft 7". As a result, the NRC has requested the Authority to provide supporting information regarding the differences between the proposed RETS and the model SRETS. Presented below is the requested information.

1. Turbine Buildina Service Water Radiation Monitorina i

The plant service water system provides a heat sink for the Reactor and Turbine Building Closed Loop Cooling Water Systems and other major components before discharging into the 1 environment. This system was designed in such a way that the l release of radioactive materials to the environment are below the limits set forth in 10 CFR 20, and will meet the l

requirements in 10 CFR 50 Appendix I and the Technical

Specifications. All of the Reactor Building service water heat loads discharges into the circulating water system via one of the service water discharge headers. Service water radiation monitor 17-RM-351 is provided at this discharge point to indicate when operational limits for the normal release of i radioactive material to the environs are being approached, and to indicate system malfunctions by detecting the presence of

radioactive material in a normally uncontaminated system.

The Turbine Building Closed Loop Cooling Water System (TBCLCW) provides cooling, thrcugh heat exchangers, to service equipment such as air compressors, turbine auxiliary systems, and pump i bearings. This TBCLCW system is cooled by the normal service 1 water system which discharges directly into the circulating water system via the other service water discharge header. No

, radiation monitor is required, since this service water pathway I system is very unlikely to contain radioactive material and components usually cooled by the TBCLCW system contain very low

' levels of radioactive material. In addition, the service water is normally operated at a higher pressure through the heat exchangers than the TBCLCW system, thus, isolating any contamination to the closed loop side. As a plant requirement.

[ weekly samples are taken of the TBCLCW system for possible l

contamination. In the past year, a high reading of 9 cpa/ml above background was reached. This is equivalent to approximately 1 x 10-5 pCi/ml. If a leak should occur fron l the closed loop side and into the service water side, an in-line radiation monitor could not detect it if the dilution i was 1 10 times. Weekly grab sample surveillance as performed is therefore an acceptable alternative to continuously monitor j the TBCLCW system.

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2. Gaseous Effluent Radiation Monitorina Indication The following is a description of the available instrument surveillance indication for gaseous effluent radiation monitors.

e A. Main stack radiation monitors (17-RM-50A,B) - Each monitor has the following indicator lights in the control room:

- DOWNSCALE (low radiation level-white)

- UPSCALE HIGH (high radiation level-amber)

- UPSCALE HI/HI (high-high radiation level-amber)

- INOP (inoperative equipment-white) i In addition, control room annuniciation is provided for the following:

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- OFF-GAS VENT PIPE HI/LO FLOW

- OFF-GAS VENT PIPE DOWNSCALE/INOP (low radiation or inoperative equipment)

- OFF-GAS VENT PIPE HIGH RAD (high radiation level)

- OFF-GAS VENT PIPE HI/HI RAD (high-high radiation level B. Ventilation Exhaust Radiation Monitoring - This t system consists of the following:

1. Turbine Building Ventilation Exhaust Radiation i Monitors (17-RIS-431,432)

. 2. Radwaste Building Ventilation Exhaust Radiation Monitors (17-RIS-458A,B)

3. Refueling Floor Ventilation Exhaust Radiation Monitors (17-RIS-456A,B)

4. Reactor Building Ventilation Exhaust Radiation Monitors (17-RIS-458A B) l The radwaste building, turbine building, and refueling

floor ventilation system radiation monitors are i identical in design. Instrumentation is provided to

! regulate and indicate sample flow rate. Local i

annunciation is provided for flow disturbance. The j following annunciators are provided:

- TURBINE BUILDING EXHAUST RAD MONITOR HI/ INOPERATIVE on local panel HV-1,

- RADWASTE BLDG EXHAUST RAD MONITOR HI/ INOPERATIVE on local panel HV-12, and

- REFUELING FLOOR EXHAUST RAD MONITOR HI/ INOPERATIVE on local panel HV-3N. l l

These annunciators are duplicated on panel 9-75 in the )

! control room. Further indications are also provided by i the control room process computer printout. )

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Tho roactor building v ntilation conitoro hnvo tha log radiation Conitors and high voltago supply units located i

in the control room on panel 09-12. The monitor front panel includes a meter calibrated in cya from 101 -106 and the following indicator lights:

- DOWNSCALE (low radiation level-white).

- UPSCALE HIGH (high radiation level-amber, i - UPSCALE HI/HI (high-high radiation level-red),and

! - INOP (inoperative equipment-white).

C. Steam Jet Air Ejector Radiation Monitors (17-RM-150A.B): <

These monitors detect off-gas radioactivity by three gamma sensitive ionization chambers. The monitors each have front panel indicator lights in the control

. room for:

- DOWNSCALE (low radiation level-white),

- UPSCALE HIGH (high radiation level-amber)

- UPSCALE HI/HI (high-high radiation level-red), and

- INOP (inoperative equipment-white).

In addition, the following annunciators are provided on ;

4 panel 09-3 in the control room: )

- OFF-GAS SAMPLE HI/LO (flow disturbance).

- OFF-GAS DOWNSCALE/INOP (low radiation level or inoperative equipment),

) - OFF-GAS HIGH RAD (high radiation level,

- OFF-GAS HI/HI RAD (high-high radiation level), and

- OFF-GAS TIMER INITIATED.

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. Automatic isolation of the off-gas outlet valve is j initiated upon any combination of the above off-gas signals.

i 3. Liould Effluent Radiation Monitorina Indication l The following is a description of the available liquid effluent radiation monitors that will ensure the required instrument surveillance:

Radwaste Discharge Radiation Monitor (17-RM-350)

Service Water Radiation Monitor (17-RM-351)

RBCLCW System Radiation Monitor (17-RM-352)

These monitors are of identical design and radiation detection is by a scintillation detector mounted in a lead shield sampling chamber. Local indication, control and alarms are provided for flow rate. Control room alarm annunciation for these monitors exist for the following conditions:

- DOWNSCALE (low radiation level-white).

- UPSCALE HIGH (high radiation level-amber,

- UPSCALE HIGH-HIGH (high-high radiation level-amber),

- INOP (inoperative equipment-white)

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l In addition, a common annunciation is provided for the three monitors in the form of a LIQUID PROCESS DOWNSCALE/INOP and LIQUID PROCESS HIGH RAD on panel 09-3 in the control room.

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4. Radioactive Gaseous Waste Sand 11na and Analysis Procram Table 3.2-1 Notations d. e. and f.

A. Twenty percent rated thermal power change in one hour i instead of 15 percent, as referenced in the SRETS, is chosen because this value currently requires the Radiation Environmental Services (RES) Department to i sample and analyze reactor coolant for gross gamma activity. Refer to Appendix A Surveillance Requirement 4.6.C.l.d. Thus, the proposed RETS is consistent with current Technical Specifications.

j The bases for sampling the reactor water is to assure

' that the limit of Specification 3.6.C of Appendix A i of 3.1 pCi/gm of dose equivalent I-131 is not ex-

caeded. This limits dose at the site boundary to 33 ren to the thyroid in the event of a steam line rupture.

l B. The main stack sample was chosen as the release point to be sampled and analyzed for start-ups, 1

shutdowns and > 20 percent rated thermal power j changes in one hour because an iodine spike would be observed at this release point since this is the direct release point for the main condenser air ejectors. t

! C. The requirement for sampling the main stack for a minimum of two consecutive days rather than seven i 4

consecutive days was provided to accomplish two things:

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1. Provide the data that the seven-day sampling i requirement was intended to supply. Wording l was included which provides for the require- ,

ment to sample for the full seven days in the >

event that the measured release did not reach steady state or show a decrease in release rate (concentration) at the stack.

2. Reduce the man-hour burden that would be required to provide sampling for seven 1

( consecutive days. In the event that a shutdown was followed by a subsequent start-up within seven days of a shutdown, i sampling would be required for up to 14

consecutive days which, as noted above, would be a significant man-hour require-i ment.

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D. With the Offgas Treatment System (including the

charcoal absorbers) in service, a 20 percent rated 4 thermal power change in one hour would not produce a measurable change of iodine at the stack i discharge. Past experience shows that LLD values were observed at the stack discharge with the offgas Treatment System (including the charcoal absorbers) in service.

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5. Offuas Treatment System Operability J

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Operating the offgas treatment system at less than 50% power j will cause the system to run so inefficiently that it will be effectively rendered inoperable. This is due to the fact that at this power level and below, only one turbine-driven feedwater i j pump is in operation which results in an increase of condenser

, air in-leakage at the feedwater pump turbine exhaust and main i

! condenser inlet. This excess air in-leakage exceeds the offgas 4 treatment system operating limitations. Main condenser air

! in-leakages, however, are decreased when the second turbine-driven feedwater pump is placed into operation. At this point, the offgas treatment system is started to ensure proper treatment of gaseous waste, and maximize recombiner efficiency.

j During normal startup after a refueling outage, reaching power level of 50% normally takes only 2 1/2 days during which a variety of tests are conducted. The plant normally does not remain at this power level and continues to increase the power level with start-up of the second turbine-driven feedwater pump.

! Since the installation of the offgas treatment system, the

, Authority has made numerous modifications in an attempt to

~j ensure proper treatment of gaseous radioactive waste.

Presently, the offgas treatment system is partially operational :

and a series of modifications is planned to restore system performance, reduce offgas leakage to the turbine building, r maximize system availability, and allow the charcoal absorbers i to be used for noble gas delay. This modification is planned to be installed in the next refueling outage, currently scheduled for February of 1985.

6. Maior Modifications to Radioactive Maste Treatment Systems Documentation of the fact that the modification has been reviewed and found acceptable by the PORC will be available at the Fit 2 Patrick plant for review.

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