ML20082G626

Proposed Change to Radiological Effluent & ETS

ML20082G626
Person / Time
Site:	Oyster Creek
Issue date:	09/26/1983
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20082G617	List: ML20082G615 ML20082G622 ML20082G626
References
NUDOCS 8311300272
Download: ML20082G626 (85)

Text

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PROPOSED CHANGE l

RADIOLOGICAL EFFLUENT AND ENVIRONMENTAL TECHNICAL SPECIFICATIONS APPENDIX A TO ,

PROVISIONAL OPERATING LICENSE OPR-16 0YSTER CREEK NUCLEAR GENERATING STATION JERSEY CENTRAL POWER AND LIGH'T COMPANY

-DOCKET NO. 50-21')

SEPTEMBER 26, 1983

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8311300272 831125 PDR ADOCK 05000219 P PDR

', ,1vl9 B. Source Check A SOURCE CHECK is the qualitative assessment of channel response when the channel sensor is exposed to a source of radioactivity.

1.26 . PROCESS CONTROL PLAN The PROCESS CONTROL PLAN shall generally describe the essential operational controls and surveillance checks for processing wet L

f radioactive waste in oroer to provide reasonable assurance of compliance with class B or C stability requirements of 10 CFR Part 61.56 (b) before disposal.

1.29 AUGMENTED OFFGAS SYSTEM (A0G)

The AUGMErTED OFFGAS SYSTEM is a system designed and installed to holdup and/or process radioactive gases from the main condenser offgas system for the purpose of reducing the radioactive material content of the gases before release to the environs.

1.30 MEMBER OF THE PUBLIC A MEMBER OF THE PUBLIC is a person who is not occupationally associated with GPU and who does not normally frequent the Oyster Creek Nuclear Generating Station site. -The category does not include contractors, contractor employees, vendors, or persons who enter the site to make deliveries to service equipment, or work on the site, or for other purposes not associated with plar.t functions.

1.31 0FFSITE DOSE CALCULATION MANUAL An 0FFSITE DOSE CALCULATION (00CM) states the methodology and parameters to be used in the calculation of radiation doses offsite due to radioactive gaseous and liquid effluents and in the calculation of radioactive gaseous and liquid effluent monitoring instrumentation alarm / trip.setpoints.

, . , ,1v32 PURGE PURGE or PURGING is. the controlled process of discharging air or gas from a confinement and replacing it with air or gas.

1.33 SITE BOUNDARY The SITE 80VHDARY is the perimeter line beyond which the land is neither owned, leased, nor otherwise controlled by GPU. For the purpose of implementing radiological effluent technical specifications, the 0FFSITE area is that land (offsite) beyond the site boundary.

1.34 SURVEILLANCE FREQUENCY A SURVEILLANCE FREQUENCY or interval may be specified for a periodic surveillance calibration, test, check, or examination according to the tabulated notation. An interval or frequency may be adjusted by

+ 25 percent.

NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> O At least once per 24_ hours W At least once per 7 days M At least once per 31 days Q At least once per 92 days S'A At least once per 184 days R At least once per 18 months S/U Before each reactor startup P Completed before each release N.A. Not Applicable

3t6 Radioactive Effluents Applicooility: Applies to the radioactive effluents of the facility.

Objective: To assure that radioactive material is not released to the environment in an uncontrolled manner and to assure that the radioactive concentrations of any material released is kept as low as is reasonably achievable and, in any event, within the limits of 10 CFR Part 20.106 and 40 CFR Part 190.102.

Specification: 3.6.A. Reactor Coolant Radioactivity The total iodine concentration in the reactor coolant shall not exceed 8.0 uCi/gm. In the event the concentration cannot be maintained at or below 8.0 pCi/gm, the reactor shall be placed in the SHUT 00WN CONDITION.

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- 3.6.B Liquid Radwasta Treatment Applicability: To liquid radwaste batches before discharge as aqueous effluent.

1. Any untt ected batch of liq ~'d radwaste shall be treated (in appropriate liquid radwaste treatment equipment) before discharge as aqueous effluent when tne radioactivity concentration, exclusive of tritium and dissolved noble gases, in the batch exceeds 0.001 uCi/ml.

2. When radioactive liquid waste is discharged without treatment and in excess of the above limit, in lieu of any other report, prepare and submit to the Commission within 30 days pursuant to Specification 6.9.3 a Special Report that includes the following information:

a. Identification of any inoperable equipment or subsystems, and the reason for the inoperability.

b. Action (s) taken to restore the inoperable equipment to OPERABLE status, and a

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

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3. Specifications 3.0.A and 3.0.8 do not apply.

,. ..- 3.6.C Radios tive Liquid Storage

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Applicability: Applies at all times to outdoor tanks used to store radioactive liquids.

1. The quantity of radioactive material, excluding tritium, noble gases, ana radionuclides having half-lives shorter than three days, contained in any of the following outdoor tanks shall not exceed 10.0 curies:

a. High Purity Warte Sample Tanks, HPT-2A and HPT-28.

b. Chemical Waste Distillate Sample Tanks, WC-T-3A and WC-T-3B, and

c. Condensate Storage Tank

d. Temporary Storage Tank (s) outside of building

2. In the event the quantity of radioactive material in any of the tanks named exceeds 10.0 curies, begin treatment without delay, continue it until the total quantity of radioactive material in the tank is 10 curies or less, and describe the reason for exceeding the limit in the next Semiannual Radioactive Effluent Release Report.

3. Specifications 3.0.A and 3.0.8 do not apply.

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,. . 3.6.D Condens*r Offgas Treatm:nt 7 .,-

Applicability: Whenever tne main condenser air ejector system is in operation except during startup or shutdown with reactor power less than 50 percent of rated. In addition, the Auguemted Offgas System need not be in operation during end of cycle coast-down periods when the system can no longer function due to low offgas flow rates.

1. Every reasonable effort shall be made to maintain and OPERATE charcoal absorbers in the Augmented Offgas System to treat radioactive gas from the main condenser

' air ejector.

2. If gaseous effluent is released without treatment for more than 30 days and either Specification 3.6.L or 3.6.M is exceeded, in lieu of any other report, submit a Special Report pursuant to Specification 6.9.3 to the NRC within 30 days which includes the following information:

a. Identification of the inoperable equipment or subsystem and the reason for inoperability; and

b. Action (s) taken to restore the inoperable equipment to OPERABLE status and to prevent a recurrence.

5 3.6.E Main Condenser Offgas Radioactivity

1. The gross radioactivity in' noble gases discharged from the main condenser air ejector shall not exceed 2.1/E Ci/sec where E is the average gamma energy (Mev per atomic transformation).

- 2. In the event Specification 3.6.E.1 is exceeded, reduce the discharge rate below the limit within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STAND 8Y~.within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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.. -s 3. 6. F . Condenser Offgas Hydrogan Concentration

1. The concentration of hydrogen in the AugmentedOffgasSystem(A0G) downstream of the recombiner during A0G operation shall not exceed 4 percent by volume.

~2. In the event the hydrogen concentration downstream of a-recombiner exceeds 4 percent by volume, the concentration shall be reduced to less than 4 percent within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

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,. . 3.6.1 Radioactivity Concentration in Liquid Efflu?.nt

1. The concentration of radioactive material, other than noble gases, in liquid effluent in the discharge canal at the doute 9 bridge (see FSAR Figure 11-2-2) shall not exceed the concentrations specified in 10 CFR Part 20, Appendix 8, Table II, Column 2.

2. The concentration of noble gases dissolved or entrained in liquid effluent in the discharge canal at the Route 9 bridge shall not exceed 2 x 10-4 microcuries/ milliliter.

3. In the event the concentration of radioactive material in liquid offluent released into the unrestricted area beyond the Route 9 bridge exceeds either the concentration limit in 3.6.I.1 or 3.6.I.2, reduce the release rate without delay to bring the concentration within the limit.

4. The provisions of Specification 6.9.2.b are not applicable.

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,. ,- 3.6.J Limit on Doss Due to Liquid Effluent

1. The dose to a MEMBER OF THE PUBLIC due to radioactive material in liquid effluents beyond the SITE BOUNDARY shall not exceed:

1.5 mrem to the total body during any calendar quarter, 5 mrem to any body organ during any calendar quarter, 3 mrem to,the total body during any calendar year, or 10 mrem to any body organ during any calendar year,

2. When the calculated dose from the release of radioactive materials in liquid effluents exceeds any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days from the end of the quarter during which the release occurred, pursuant to Specification 6.9.3, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken and/or will be taken.

3. The provisions of Specifications 3.0.A and 3.0.8 are not applicable.

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- 3.6.K Dose Rate DuS to Gascous Effluent

1. The dose equ valent rate beyond the SITE BOUNDARY (see FSAR Figure 11-2-2) due to radioactive noble gas in gaseous effluent shall'not exceed 500 mrem / year to the total body or 3000 mrem / year to the skin.

2. The dose equivalent rate beyond the SITE BOUNDARY due to I-131, I-133, and to radioactive material in particulate from having half-lives of 8 days or more in gaseous effluents shall not exceed 1500 mrem / year to any body organ.

3. In the event the dose equivalent rate exceeds any of the limits in 3.6.K.1 or 3.6.K.2, decreases the' release rate without delay to comply with the ilmit. If the gaseous effluent release rate cannot be-reduced to meet the limits, the reactor shall be in at least H0T STANDBY within 48

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hours unless corrective actions have been completed and the release rate restored to below the limit.

4. The provisions of Specification 6.9.2.0 do not apply.

- 3.6.L Air Dose to Noble Gas in Gaseous Effluent

1. The air dose beyond the SITE BOUNDARY (see -

FSAR Figure II-2-2) due to noble gas released in gaseous effluent shall not exceed:

5 mrad / calendar quarter due to gamma radiation, 10 mrad / calendar quarter due to beta radiation, 10 mrad / calendar year due to gamma radiation, or 20 mrad / calendar year cue to beta radiation.

2. If the calculated air dose to noble gas released in gaseous effluent exceeds any limit in Specification 3.6.L.1, prepare and submit a Special deport .to the Comniission which identifies the cause(s) for exceeding the limit and describes the corrective action taken, The Special Report shall:oe pursuant to Specification 6.9.3, shall be in lieu of any other report, and shall be submitted to the Commission within 30 days l from the end of the quarter during which the t.

release occurred.

3. The provisions of Specifications 3.0.A and I

3.0.8 do not apply.

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, 3.6.M Dose Ous to Radiciodine and Particulates in Gaseous Effluent

1. The dose to a MEM8ER OF THE PUBLIC from iodine-131, iodine-133, and from radionuclides in particulate form having half-lives of 8 days or more gaseous effluents beyond the SITE 80VNDARY shall not exceed 7.5 mrem to any body organ per calencar quarter or 15 mrem to any body organ per calendar year.

2. When the calculated dose from I-131, I-133, and from radionuclides in particulate form having half-lives of 8 days or more in gaseous effluent exceeds any limit in Specification 3.6.M.1, prepare and submit a Special Report to the Commiss' ion whicn identifies the cause(s) for exceeding the limit and describes the corrective action taken. The Special Report shall be pursuant to Specification 6.9.3, shall be in lieu of any other report, and shall be submitted to the Commission within 30 oays from the end of the quarter during which the release occurred.

3. The provisions of Specifications 3.0.A and 3.0.8 do not apply.

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.' 3.6 N Annual Total Dose Due to Radioactive Efflu:nts

1. The annual cose to a MEMBER OF THE PUBLIC due to radiation and radioactive material in effluents from the OCNGS shall not exceed 75 mrem to his thyroid or 25 mrem to his total body or to any other organ.

2. In the event the calculated dose due to radioactive material released in liquid or gaseous effluent exceeds twice the limits of Specification 3.6.J.1, 3.6.L.1, or 3.6.M.1, perfonn an assessment of compliance with Specification 3.6.n.1 in accordance with methodology in the 00CM.

3. In the event an assessment shows Specification 3.6.N.1 to have been exceeded, prepare and submit a Special Report to the Commission within 30 days, pursuant to Specification.6.9.3 and in lieu of any other report. The report shall include information specified in 10 CFR 20.405(c).

If the condition causing the limit (s) to be exceeded nas not been corrected, the Special Report may also state a request for a variance in accordance with the provisions of 40 CFR Part 190. In that event, the request is timely and a variance is granted until NRC action on the request is complete.

4. Tne provisions of Specification 3.0.a ana 3.0.8 do not apply.

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BASES

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Basis: 3.6.A teprimarycoolantradioactivityconcentrationlimitof8.g Ci total iodine per gram of water calculated is based on a steamline break accident which is isolated in 10.5 seconds.

For this accident analysis, all the iodine in the mass of coolant released in this time period is assumed to be released to the atmosphere at the top of the turbine building (30 meters). By limiting the thyroid dose at the site boundary to a maximum of 30 Rem, the tolerable iodine concentration in the primary coolant is back-calculated assuming fumigation meterology, Pasquill Type F, and 1 m/sec. wind speed. We iodine concentration in the primary coolant resulting from this analysis is 8.4 uCi/gm.

Basis: 3.6.B m is specification implements the requirements of 10 CFR Part 50.36a; 10 CFR Part 50, Appendix A, criterion 60, and 10 CFR part 50, Appendix I,Section II.D.  % e retention of radioactive liquids on site is consistent with maintaining radioactive discharge as low as reasonably achievable.

he radwaste liquid effluent from the Oyster Creek Station will be controlled on a batch basis with each batch being processed by a method appropriate for the quality and quantity of materials present. We operability of the liquid radwaste treatment system ensures that the appropriate portions of the system will be available for use whenever liquid waste requires treatment before release to the environment, i

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Th:s0 batch s in which the radioactivity concentrations are sufficiently low to allow release to the discharge' canal are diluted with condenser circulating water and/or dilution water. The radioactive liquids in batches discharged from the liquid radwaste treatment system are sampled and analyzed for radioactivity prior to release to the discharge canal, thus providing means for obtaining information on effluent to be released so that appropriate release rates can be established.

Specification 3.6.8 implements 10 CFR Part 50 Appendix I provisions for-cost-beneficial treatment of radioactive liquid waste before release in effluent.

- Basis: 3.6.C Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix 8, Table II, Column 2 in the canal at the Route 9 bridge.

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Rstaining radioactive liquids on-site in order to permit systematic and appropriate processing is consistent with maintaining radioactive discharges to the enviror. ment as low as practicable. Limiting the contents of each outside tank to 10 curies or less assures that even if the contents of a tank were released onto the ground and drained into the discharge canal, the potential dose to a memoer of the public is estimated to be less than 1 percent of the 500 mrem / year limit t'o the. total body of a member of the public only 1 percent of the corresponding 1500 mrem / year standard for a single organ.

In the highly unlikely event that every outside tank

. named in Specification 3.6.C were to contain 10 curies and the contents of all were to spill into the

- discharge canal, the potential dose to a menber of the public is estimated to be only about 2 percent of the 500 mrem / year limit to the total body and about o percent of the corresponding 1500 mrem / year standard.

Basis: 3.6.0 The operability of the AUGMENTED OFFGas SYdTEM (A0G) charcoal absorcer ensures that they will be available for use whenever main condenser offgases require treatment prior to release to the environment.

, e The appropriate portions of this system provide reasonable assurance that the releases or radioactive materials in gaseous effluents-will be kept "as low as is reasonably achievable". A Special Report is required in the event the Augmented.0ffgas System charcoal absorber is not OPERATED and a concentration or dose exceeds a relevant limit offsite.

Basis: 3.6.E Some radioactive material is released from the plant under controlled conditions as part of the normal operation of the facility. Other radioactive material s

not normally intended for release could be inadvertently released in the event of an accident.

l Therefore, limits in 10 CFR Part 20 apply to releases during nonnal operation and limits in 10 CFR Part 100 app.;' to accidental releases.

Radioactive gases from the reactor pass through the steam lines to the turbine and then to the main condenser where they are extracted by the air ejector, passed through holdup piping and released via the plant stack. Radioactive materials release limits for the plant stack have been calculated using meterological data from a 4Uu ft. tower at the plant site. The analysis of these on-site meterological data shows that a release of radioactive gases after 30 minutes holdup in the offgas system of 0.3 Ci/sec.,

which would not result in a whole body radiation dose exceeding the 10 Cr3 20 v31ue of 0.5 rem per year.

1- The Holland plume rise model with no correction factor was used in the calculation of the effect of momentum and buoyancy of a continuously emitted plume.

~ Independent dose calculations for several locations offsite were made by the AEC staff from onsite meteorological data developed by the licensee and diffusion _ assumptions appropriate to the site. The procedure followed is described in Section 7-5.2.5 of

" Meteorology and Atomic Energy - 1968", equation 7.63 being used. The results of these calculations were equivalent to those generated by the licensee provided the average gamma energy per_ disintegration for the assumed noble gas mixture with a 30 minute holdup is 0.7 MeV per disintegration. Based on these calculations, a maximum release rate limit of gross activity, except for iodines and particulates with half lives longer than eight days, in the amount of

-0.21/E curies per second will not result in offsite annual doses in excess of the limits specified in 10 CFR Part 20. The E determination need consider only the average gamma energy per disintergration since the controlling whole body is due to the cloud passage over the receptor and not cloud submersion, in which the beta dose could be adoitive.

Basis: ~3.6.F Tha purpose of Sp:cification 3.b.F is to require that

'the concentration of potentially explosive gas mixtures in the Augmented Offgas System be maintainea below the flammability limit of hydrogen in air, although the AUG is explosion resistant.

Specification 3.6.F applies-to the hydrogen concentration downstream of a recombiner during AUG operation. The A0G has redundant recombiners so that the recombiner in use can De isolated and purged with air in the event hydrogen in it exceeds the specified limit.

Basis: 3.6.I The purpose of Specification 3.6.I is to require that concentrations of radioactive material in aqueous effluents to unrestricted areas comply with 10 CFR Part 20.106. The concentration limit for dissolved or entrained noble gas is based on assumed exposure by immersion in water containing Xe-135 (assumed to be the critical radioactive noble gas). The concentration limit of noble gases is applied inaependently of the limit for other radionuclides because the exposure pathway is separate.

, Basis: 3.6.J Tha purpose of Specification 3.6.J is to require compliance with 10 CFR Part 50 Appendix I, Section IV.A to assure that radioactive material in liquid effluent is kept as low as is reasonably achievable ano to permit operating flexibility under unusual operating conditions.

Basis: 3.6.K The purpose of Specification 3.6.K is to require that concentrations of radioactive material in airborne effluents to unrestricted areas comply with 10 CFR

'Part 20.106. The occupancy of a Member of the Public who may from time to time be within the Site Boundary is taken to be sufficiently low to compensate for any increase in atmospheric concentration within the Site, thereby causing the exposure of those Members of the Public to be less than the equivalent annual limit on radiation exposure to a Member of the Public incurred offsite.

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f-i Basis: 3.6.L and 3.6.M The purpose of Specification.3.6.L and 3.6.M islto require compliance with 10 CFR Part 50 Appendix-I,Section IV.A^and to provide operating flexibility under unusual

.. operating conditions as permitted in 10 CFR Part 50.36a.

Assessment of compliance is implemented by calculational methods specified in-the 00CM provided by the Surveillance Requirements. The 00CM methodology provides for assessing compliance with dose limits at or beyond the Site Boundary

-based on either historical average atmospheric conditions or conditions averaged over the period of interest.

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, 3.14 Solid Radioactive Waste Applicability: Processing wet radioactive waste destined for disposal by burial in land as class B or C waste.

Objective: To provide reasonable assurance that the applicable waste satisfies stability requirements for classes B and C wastes stated in 10 CFR Part 61.56(b) before disposal.

Specification: 3.14.A Wet radioactive waste destined for disposal by land burial as class B or C waste shall be processed and/or containeo in accordance with a Process Control Plan to meet appropriate waste stability characteristics required by 10 CFR Part 61.5b before being shipped to a disposal facility.

3.14.8 In the event waste processing and/or placement into a container for the purpose of meeting appropriate class b or C stability requirements of 10 CFR Part 61.56 is performed by a contractor, GPU shall verify that the contractor has an NRC approved Process Control Plan.

3.14.C The provisions of 3.0.A, 3.0.8 and 6.9.2.b do not apply.

, , Basis: 3.14 10 CFR Part 61.55 defines classes B and C radioactive wastes which, among others, must meet certain requirements on waste form to insure stability after disposal. 10 CFR 61.56 states the requirements which apply to characteristics of radioactive waste being disposed of by land burial. Specification 3.14 and 4.14 apply essential operational controls and surveillance checks to processing wet radioactive waste'and/or placing it~into a high integrity container in order to provide reasonable assurance that it. satisfies the class B and C stability requirements 10 CFR Part 61.56(b) before disposal.

Regulations dealing with packaging and transport of radioactive materials to a licensed burial site are separately applicable.

It is not required that the processing and/or

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placement of wastes into a high integrity container be done by GPU. In the event a contractor performs the service, only item B of Specification 3.14 applies.

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. ,- 3.15. Radioactiva Effluent Monitoring Instrumentation Applicability: Applies to instrumentation whose function is to monitor aqueous and airborne radioactive effluents from the Station.

Objective: To assure that instrumentation to monitor radioactive effluents is OPERABLE when effluent is discharged or that means of measuring effluent is provided.

Specification: 3.15.A Liquid Effluent Instrumentation

1. The radioactive liquid effluent monitoring channels listed in Table 3.15.1 shall be OPERABLE with their alarm / trip setpoints set to initiate alarm / trip in the event the limit of Specification 3.6.I.1 is exceeded.

2. The alarm or trip setpoint of these channels shall be detennined and set in accordance with the method described in the ODCM.

3. When a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint is less conservative than required by the above specification, without delay suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperaole, or change the'setpoint so it is acceptably conservative, or provida for manual initiation of the alarm / trip function (s).

4. When less than the minimum number of radioactive liquid effluent monitoring instrumentation channels are OPERABLE, take the ACTION shown in Table 3.15.1.

Make every reasonaole effort to restore the instrument to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

The Provisions of Specifications 3.0.A, 3.0.8, and 6.9.2.b are not applicable.

3.15.8 Gaseous Effluent Instrumentation

1. The radioactive gaseous effluent monitoring channels listed in Taole 3.15.2 shall be OPERABLE with their alarm / trip setpoint set to cause automatic alarm in the event a limit of Specification 3.6.K.1 is exceeded.

2. The Condenser Air Ejector Offgas Radioactivity Monitor shall be OPERABLE as stated in Table 3.15.2. Its trip setpoint shall be set to initiate offgas isolation in the event Specification 3.6.E.1 is exceeded.

,- TABLE 3.15.1

-RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum Instrument Channels Applicability Action Operable

1. GROSS RADI0 ACTIVITY MONITORS

a. Liquid Raawaste Effluent Line 1 b 110

b. Reactor Building Service Water System Effluent Line 1 b 112

c. Turbine Building Sump No.1-5 1 b 112

2. FLOW MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line 1 b 113 e

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Table 3.15.1 Notations

a. Instrument channels shall be OPERA 8LE and in service as indicated except that

.a channel ~ may be taken out of service for the purpose of a check, calibration, test, or maintenance without declaring the channel to be inoperable.

b. During releases via this pathway.

c. At-all times.

ACTION 110 with no channel OPERABLE, effluent may be releaseo provided 't hat before initiating a release:

1. at least two samples are analyzed in accordance with Specification 4.6.I.1, and

2. a technically qualified Staff member determines the acceptable release rate and proper discharge valving and another technically qualified Staff member verifies that the release rate and discharge valving are acceptable.

Otherwise, suspend release of racioactive effluent via this pathway.

ACTION 112 With not channel OPERA 8LE, effluent releases via this pathway may continue provided that, at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> during the release, grab samples are collected ano analyzed for gross radioactivity (beta or gamma) at a limit of detection of at least 10-0 microcuries/ml. 9 ACTION 113 With no channel OPERA 8LE effluent releases via the affected pathway may continue provided the flow is estimated with the pump curve at least once per batch during a release.

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i' 'r TABLE'3.15.2-1' ,

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RADIOACTIVE GASEOUS' EFFLUENT. MONITORING INSTRUMElffATION ~

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. Instrument: Channels. Essential Applicability Action '

Operable Function Main Condenser Offgas Treatment System Hydrogen 2 Monitor hydrogen c 125-

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concentration-

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Monitor

2. Stack' Monitoring System (RAGEMS).

1- . Monitor activity. b,e 124

a. Radioactive Noble Gas Monitor concentration, alarm 1- Cbllect sanple b,e 127-

' b. Iodine Sanpler

. b, e 127-

c. Particulate Sanpler 1 Collect sanple

d. Effluent Flow Measuring Device- 1 Measure air flow b cl22

i. Measure air flow b 128~

e. Sanpler Flow Measuring Device 1

3. Turbine Building Ventilation Monitoring System (RAGEMS) 123 1 . Monitor activity b

[ a. Radioactive Noble Gas Monitor concentration 1 . Collect sanple b. '127

b. Iodine Sanpler

c. Particulate Sanpler 1 ' Collect sanple b '127 1 Measure air flow b 122

d. Effluent Flow Measuring Device

.1 Measure air flow- b 128

e. Sanpler Flow Measuring Device

4. Offgas Building Exhaust Ventilation Monitoring System (RO76) b 123 1 Monitor activity

a. Radioactive Noble Gas Monitor concentration 1 Collect sanple b 127

b. Iodine Sanpler 1 Collect sanple b 127.

c. Particulate Sanpler Measure air flow b 122

.d. Effluent Flow Measuring Device 1 b 123 -

e. Sanpler Flow Measuring Device 1 Measure air flow l

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TABLE 3.15.2-2 PRUfBCTIVE INS 1HUBENFATION HBQUIREMENTS >

Reactor Modes Min. No. of Min No. of in Which Function Operable or Operable Must Be Operable Operating Cannels Per (Tripped) Trip Operable Action Function Trip Setting Shutdown Refuel Startup Run Systems Trip Systemas Required *

1. Iow-Iow-Iow _4'8' above X (V) X (V) X (V) .X 2 2 See note h Reactor Water top of Ievel active fuel

2. Ac Voltage NA X (V) X 2 2 Prevent auto depressurization on loss of AC power. See note i Isolation Condenser Isolation

1. High Flow steam _20 peig P X (S) X (S) X X 2 2 Isolate Affected isolation Line condenser comply with Spec.

3.8

2. Hign Flow Con- 27" P 1120 X (S) X (S) X X 2 2 densate Line Offgas System Isolation

1. High Radiation 2.1 ciE X (S) X (S) X X 1(c) 1(c) Note d In Offgas Line E Sec Reactor Building Isoli.nion and Standby Gas Treatment System Initiation

1. High Radiation - 100 mr/Hr X (W) X (W) X 1 1 Isolate Reactor Reactor Bldg. Bldg. & Initiate Operation Floor Standby Gns Treatment or Manuel Surveillance for not more ttan 24

2. Reactor Bldg. _ 17 mr/Hr X (W) X (W) X X 1 1 hours (total for Ventilation all Ins truments Exhaust under .1) in any 30 day period.

3. High Drywell _ 2 psig X (W) X (W) X X 1 (k) 2 (k)

Pressure 7'2" above X X X X 1 2

4. Iow WaterIow Reactor - top of active Level fuel

TABLE 3.15.2 NOTATIONS

$ a. Channels shall be OPERABLE and in service as indicated except that a channel may be taken out of service for the purpose of a check, calibration, test, maintenance or sample media change without declaring the channel to be inoperable.

b. During releases via this pathway.

e. Monitor / sampler or an alternate shall be OPERABLE to monitor / sample Stack effluent whenever the drywell is being purged.

ACTION 122 With no channel OPERABLE, effluent releases via this pathway may continue provided the flow rate is estimated and recorded whenever the exhaust fan combination in this system is changed.

ACTION 123 Hith no channel OPERABLE, effluent releases via this pathway may continue provided a grab sample is taken at least once per 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> an is analyzed for gross radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter or provided an alternate monitoring system with local display is utilized and the measurement is observed and recorded at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

ACTION 124 With no channel OPERABLE, effluent releases via this pathway may continue provided a grab sample is taken at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and analyzed for gross radioactivity with 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or provided an alternate monitoring system with local display is utilized. Drywell purge is permitted only when the radioactive noble gas monitor is operating.

ACTION 125 With one channel OPERABLE, operation of the main condenser offgas treatment system may continue provided a recombiner temperature sensing instrument is operable. When only one of the types of instruments, i.e., hydrogen monitor or temperature monitor, is operable, the offgas treatment system may be operated provided a gas sample is collected at least once per day and is analyzed for hydrogen within four hours.

~

ACTION 127 Hith no channel OPERABLE, effluent releases via this pathway may continue provided the required sampling is initiated with auxiliary sampilng equipment as soon as reasonable after discovery of inoperable primary sampler (s). The auxiliary sampling and analysis shall be performed in accordance with Table 4.6.2.

ACTION 128 With no channel OPERABLE, effluent releases via the sampled pathway may continue provided the sampler air flow is estimated and recorded at least once per day.

4

4 TABLE 3.15.2 (Cont'd)

1. These functions not required to be operable when the ADS is not required to be operable.

j. These functions must be operable only when irradiated fuel is in the fuel pool or reactor vessel and secondary containment integrity is required by Specification 3.5.B.

k. The number of operable channels may be reduced to 2 per Specification 3.9-E and F.

1. The bypass function to permit scram reset in the shutdown or refuel mode with control rod block must be operable in this mode.

m. Pump circuit breakers will be tripped in 10 seconds 15% during a LOCA by relays SK7A.and SK8A.

n. Pump circuit breakers will trip instantaneously during a LOCA.

o. Instrument shall be operable during main condenser air ejector operation except that a channel may be taken out of service for the purpose of a check, calibration, test, or maintenance without declaring it to be inoperable.

p. With no channel OPERABLE, main condenser offgas may be released to the environment for as long as 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided:

(a) The offgas treatment system charcoal absorbers are not bypassed, and (b) the stack radioactive noble gas monitor is OPERABLE.

Otherwise, be in at least HOT STANDBY within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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TABLE 3.15.2 (Cont'd)

Action required when minimum conditions for operation are not satisfied.

Also permissible to trip inoperable trip system. When necessary to conduct tests and calibrations, one channel may be made inoperable for up-to one hour per month without tripping its trip system.

See Specification 2.3 for Limiting Safety Sytem Settings.

NOTES:

a. Permissible to bypass, with control rod block,-for reactor protection system react in refuel mode.

~ b. Permissible to bypass below 600-psig in refuel and startup modes.

c. One (1) APRM in each operable trip system may be bypassed or inoperable provided the requirements of Specification 3.1.C and 3.10.0 are satisfied. Two APRM's in the same quadrant shall not be concurrently bypassed except as noted below or permitted by note.

Any or,e APRM may be removed from service for up to one hour for test or calibration.without inserting trips in its trip system only if the remaining operable APRM's meet the requirements of Specification 3.1.B.1 and no' control rods are moved outward during the calibration or test.

~During this short period, the requirements of Specifications 3.1.B.2, 3.1.C and 3.10.D need not be met.

d. The (IRM) shall-_be inserted and operable until the APRM's are operable and reading at least 2/150 full scale.

e. Offgas system isolation trip set at _ 2.1/E C1/sec where E = average gamma energy from noble gas in offgas, f .' Unless SRM chambers are fully inserted.

g. Not applicable when IRM on lowest range.

h. One instrument channel in each trip system may be inoperable provided the circuit which it operates in the trip system is placed in a simulated tripped condition. If repairs cannot be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> the reactor shall be placed in the cold shutdown condition. If more than one instrument channel in any trip system becomes inoperable the reactor shall be placed in the cold shutdown condition. Relief valve controllers shall not be bypassed for_more than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (total time for all controllers) in any 30-day period and only one relief valve controller may be bypassed at a time.

TABLE 3.15.3 -

INSTRUMENT CALIBRATION SCHEDULE Instrument Channel Check Calibrate Igit Remarks (Anolier to Test and Calibration)

ApRM Scram Trips Note 2 1/wk I/wk Using built-in calibration equipment dur-ing power operation APRM Rod Blocks Note 2 1/3 mo 1/mo Upscale and downscale

a. High Radiation on Main 1/s 1/3 mo 1/wk Using built-in calibration equipment dur-Steamline ing power operation

b. Sensors for 13 (a) N/A Each Refueling N/A Using external radiation source Outage High Radiation in Reactor Building

' Operating Floor 1/s 1/3 mo. 1/wk. Using gansna source for calibration Ventilation Exhaust 1/s 1/3 mo. 1/wk. Using gansna source for calibration High Radiation on Air 1/s Channel Check Ejector Off-Gas 1/m Source Check Each refueling Calibration according to established '

outage station calibration procedures.

Each refueling Note 2.

outage l IRM Level N/A Each Shutdown N/A During approach to shutdown only IRM Scram Using built-in calibration equipment IRM Blocks N/A Prior to Start- Prior to Start- Upscale and downscale up & Shutdown up & Shutdown .

Condenser Low Vacuum N/A Each Refueling Each Refueling outage outage i

Calibrate prior to startup and normal shutdown and thereafter check 1/s and test 1/wk until no longer required.

Leaend' N/A = Not applicable; 1/s = Once per shift; 1/d = Once per day; 1/3d = Once per 3 days; 1/wk = Once per week; 4 1/2 mo. = Once every 3 months.

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.. BASES Basis: 3.1.5.A The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive e' materials in liquid effluents during actual or potential releases of liquid effluents. The alarm / trip setpoints for these instruments are calculated and adjusted in accordance with the methodology and parameters in the 00CM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part

20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50. The purpose of tank level

' indicating devices is to assure the detection and control of leaks that is not controlled could potentially result in the transport of radioactive materials to UNRESTRICTED AREAS. Radioactivity monitors on the liquid radwaste effluent line and in the Turbine Building Sump No. 1-5 initiate a trip to stop effluent discharge when the trip setpoint is exceeded.

. ' Basis: 3.15.8 The radioactive gaseous effluent instrumentation is provided to

. monitor.and control, as applicable, the releases of radioactive materials in gaseous effluents during releases of gaseous

' effluents. The alarm / trip setpoints for these instruments are calculated and adjusted in accordance with the methodology and

. parameters in the ODCM-to ensure that the alarm / trip will occur

. prior to exceeding the limits of-10 CFR Part 20. This ir;trumentation.also incluaes provisions for monitoring the concentrations of potentially explosive gas mixtures in the offgas

~ system. .The OPERABILITY and use of this instrumentation is consistent with the-requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The offgas hydrogen monitor and the radioactive gas monitors for~the condenser air e

1 ejector.offgas, the stack effluent, and the offgas building exhaust ventilation have alarms which report in the reactor control room.

The Stack and the Turbine Building Exhaust Ventilation effluent air are monitored by a radioactive gaseous effluent monitoring system (RAGEMS). It can sample, and analyze effluent for radioac'tive particulates, iodine, and noble gases. It can measure the concentration of radioactive noble gases and can identify the principal noble gas radionuclides. In the event the capability to identify. noble gas radionuclides on-line becomes inoperable, a grab sample of the effluent air will be taken at least once per month and analyzed for the principal noble gas radionuclides.

1

Sase] 3.15.B Cont'd

. Tne gross gansna activity concentration of noble gas in Stack effluent is displayed in the reactor control room. @at channel also causes an alarm in the reactor control room in the event a high activity concentration setpoint is exceeded. Iow flow of sampled Stack effluent through the RAGEMS would also cause an alarm in the reactor control room.

%ese and other data collected by the RAGEMS are displayed separately via a control conputer. In the event the control computer and or control room display fail to function or are voluntarily taken out of service, sampling particulates and iodine can be continued; the noble gas activity concentration, the sample flow, and the sampled stream flow (stack and turbine building vent) can be observed at a display located near the monitoring instrument (in which case the affected channel continues to serve its essential function and remains OPERABLE.

If the noble gas activity concentration display and the associated alarm become inoperable in either the reactor control room or in the local display monitor, then OCNGS will perform the appropriate action according to Table 3.15.2 or will provide an auxiliary monitoring system. @is permits OCNGS to retain the GE instrumentation as an alternate noble gas monitor or to substitute another noble gas monitor, particulate sampler, and iodine sampler in the event the RAGEMS and/or the NaI noble gas monitor which displays in the reactor control room becomes inoperable. We alternate sampling and/or monitoring system would be subject to the requirements stated in Specifications 3.15.B and.4.15.B.

Bases 3.15.8 Cont'd Purging the drywell to purify its atmosphere may discharge most of i the air and gases in a brief time. Hence, the drywell is purged only when the radioactive noble gas monitor in the stack monitoring system is operating in order to ensure measurement of radioactive gases discharged. Frequently, the drywell is vented to control its pressure. But since the release rate is comparatively small, the effluent is monitored as usual and the extra requirement in Table 3.15.2 Action 124 that is applied during purging is not imposed during drywell venting.

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4.6 RADIOACTIVE EFFLUENT Applicability: Applies to monitoring of gaseous and liquid radioactive-effluents of the Station during release of effluents via the monitored pathway (s). Each Surveillance Requirement applies whenever the corresponding Specification is applicable unless otherwise stated in an individual Surveillance Requirement.

Surveillance Requirements do not have to be performed on '

inoperable equipment.

Objective To measure radioactive effluents adequately to verify that radioactive effluents are as low as is reasonably achievable and within the limit of 10 CFR Part 20.106.

Specification: 4.6.A Reactor Coolant Applicability: During reactor Power Operation.

1. Reactor coolant shall be sampled and analyzed at least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to determine the radioactive iodine concentration.

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4.6.D Main Condenser Offgas Treatment

1. Operation of the Offgas System charcoal absorbers shall be verified by verifying the A0G bypass valve alignment or alignment indication.

4. 6. E - Main Condenser Offgas Radioactivity

1. The gross radioactivity in fission gases oischarged from the main condenser air ejector shall be measured by sampling and analyzing the gases

a. at least once per month, and

b. When the reactor is operating at more than 25 percent of rated power, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after an increase in the fission-gas release via the air ejector of more than 50 percent, as indicated by the Condenser Air Ejector Offgas Radioactivity Monitor after factoring out increase (s) due to change (s) in thermal power level.

s - .4.6.F Cond=nser Offgas Hydrogen Concentration-The concentration of hydrogen in offgases downstream of the recombiner in the Offgas System shall be monitored with hydrogen monitoring instrumentation as described in Table 3.15.2.

4.6.1 _ Radioactivity Concentration in Liquid Effluent

' l'. Radioactive liquid wastes shall be sampled and

, analyzed according to the sampling and analysis

- program in Table 4.6-1.

Alternately, pre-release analysis of batch (es) of radioactive liquid waste may be by gross beta or gamma counting provided a maximum concentration limit of 1 x 10-7 uCi/ml~1n the discharge canal at the Route 9 bridge is applied.

2. The results of pre-release analyses and calculation methods in the ODCM shall be used to establish i

setpoints for controlling release of each liquid radwaste batch for the purpose of assuring compliance with Specification 3.6.I.

-% e- - ,, r ,w w.-ss e---, --r------- sve- yr- -----= ' ~ = - --- , - - -- r-r -y'= 'r

.,. . 3. The alarm or trip satpoint of these

'. channels shall be determined and set in accordance with the method described in the ODCM..

4. When a radioactive effluent monitoring instrumentation channel alarm / trip setpoint is less conservative than required by specification, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel or declare the channel inoperable or change the setpoint to it is acceptable conservative, or provide for manual initiation of the alann/ trip function (s).

5. When less than the minimum number of radioactive gaseous monitoring instrumentation channels are OPERABLE, take the ACTION shown in Table 3.15.2.

Make every reasonable effort to restore the instrument to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

6. The Provisions of Specifications 3.0.A, 3.0.8, and 6.9.2.b are not applicable.

,. 4.6.J Dose Dun to Liquid Effluent

~

'. An assessment shall be performed in accordance with the ODCM at least once a month to determine compliance with Specification 3.6.J.

4.6.K Dose Rate Due to Gaseous' Effluent Radioactive noble gaseous effluent shall be monitored in accordance with Specification 3.15.B.

Radioactive noble gas monitors named in Table 3.15.2 shall be set to cause automatic alarm when the monitor setpoint, determined as specified in the ODCM, is exceeded.

4.6.L Air Dose Due to Noble Gas '

An assessment shall be performed in accordance with the ODCM at least once every month to verify that the cumulative air dose due to noble gas released in gaseous effluent does not exceed any limit in 3.6.L.l.

4.6.M Dose Due to Radioiodine and Particulates in Gaseous Effluent An assessment shall be performed in accordance with the ODCM at least once every month to verify that the cumulative dose from I-131, I-133, and radionuclides in particulate form with half-lives of 8 days or more released in gaseous effluent does not exceed any limit in Specification 3.6.M.1.

4.6.N Annual 'Ibtal Dose Due to Radioactive Effluents

1. 'Ibe cumulative dose to a Member of the Public offsite contributed by liquid and gaseous effluents shall be evaluated in accordance with the methodology and parameters in the ODCM at least once per year.

, ... TABLE 4.6.1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Lower Limit Minimum of Detectiona Liquid Release Sampling Analysis Type of Activity (LLD)

Type Frequency Frequency Analysis (uCi/ml)

A. Batch Waste P Pc Principal Gamma Release Tanks Each Batchb Each Batch Emitters 1 x 10-6 I-131 1 x 10-6 P M Dissolved and 1 x 10-6 One Batch /Mb Entrained Gases (GammaEmitters)

P M H-3 1 x 10-5 Each Batchb Composited Gross Alpha 1 x 10-7 P Q Sr-89, Sr-90 5 x 10-8 Each Batchb Composited I

- B. Reactor Building W W Principal Gamma 1 x 10-6 4

Service Water Grab Sample Emitters Effluent and Turbine Bldg. I-131 1 x 10-6 Sump No. 1-5 -

c (note f) M H-3 1 x 10-5 i

Composite 9 Gross Alpha 1 x 10-7 i

l (note f) Q Sr-89, Sr-90 5 x 10-8 l-Composite 9 Fe-55 l'x 10-6 l^

TABLE 4.6.1 NOTATIONS

a. The LLD is defined, for purposes 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 percent probability-with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation: ~

LLD = 4.66 Su E

• V* 2.22 x 106 g exp (-A at) where LLD is the lower limit of detection as defined above (microcuries per unit mass or volume)

Sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute)

E is the counting efficiency (counts per disintegration)

V is the sample size (units of mass or volume) 2.22 x 106 is the number if 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 is the elapsed time between the end of sample collection and the time of counting.

Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions with typical values of E, V, Y, and A t for the radionuclides Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 Occasionally background fluctuations, unavoidably small sample sizes, interfering radionuclides, or other uncontrollable circumstances may render these LLDs unachievable.

,, TABLE 4.6.1 NOTATIONS When calculating the LLD for a radionuclide determined oy gamma ray spectrometry, the background may include the typical contributions of other radionuclides normally present in the samples. TheLbackground count rate of a Ge(Li). detector is determined from background counts that are determined to be within the full width of the specific energy band used for the quantitative l

analysis for that radionuclide.  !

l The principal gamma emitters.for which the LLD specification will apply are '

exclusively the following racionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, )

Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144. .This list ooes not mean that only l these nuclides are to be detected and reported. Other peaks'which are  ;

measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLO for the analysis should not be reported.

b. A batch release is the discharge of liquid wastes of a discrete volume.

Before sampling for qalysis, each batch should be thoroughly mixed,

c. In the event a gross radioactivity analysis is performed in lieu of an isotopic analysis before a batch is discharged, a sample shall be analyzed for principal gamma emitters afterward.

d.- A. composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released.

e. Analysis may be performed after release. ,

f. In the event a grab sample contains more than 1 x 10-6.uCi/ml of:I-131 and principal gamma emitters or in the event the effluent'rcJioactivity monitor indicates more than 1 x 10-6 uCi/ml radioactivity in effluent, as

. applicable, sanple Reactor Building Service Water Effluent daily or sample Turbine Building Sump N.1-5 each discharge until analysis confinns the activity' concentration in the effluent does not exceed 1 x 10-6 uCi/ml.

g. A composite sanple is produced by comoining grab samples, each having a defined volume, collected routinely from the sump or stream being sampled.

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_ _ - - __.';. __ i

Print w/LM 018 TABLE 4.6.2- , , . ,

RADI0 ACTIVE GASE0tt.; WASTE SAMLING AND ANALYSIS PROGRAM ,

Minimum Lower Limit of Sampling Analysis Type of Detectiona (LLD)

Gaseous Release Twoe Fremmocv Frem =qcv Activity Analysis. ( C1/all Q Q H-3 'l x 10-s Stack Grab Sample' M M Princ! pal Gaauna Emitters e 1 x 10-*

Grab Sample (noble cases)

Stack Turbine Building c.(,f Exhaust Vent; Offgas Building Vent; Stack; and Turbine Butiding Exhaust Vent W I-131 '1 x 10-i2 Continuous' Charcoal Sample I-133 1 x 10-

! W I Continuous' Particulate Principal Gamma Emitters

• 1 x 10-^

Sample (particulates)

ME Continuous' Compostte Gross Alpha 1 x 10-'8 Particulate

_.Samole Q'

Continuous Composite Sr-89. Sr-90 1 x 10-

Particulate Sangle Noble Gas Continuous Monitor Noble Gases Gaauna Radioactivity 1 x 10-s

l TABLE 4.6.2 NOTATIONS

a. The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will oe detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement, which may include radiochemical separation:

D LLD = 4.66 S E V 2.22 x 106 Y exp (-A at)

Where LLD is the lower limit of detection as defined above (microcuries per unit mass or volume)

Sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute)

E is the counting efficiency (counts per disintegration)

V is the sample size (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 is the elapsed time between the end of sample collection and the time of counting.

Analyses shall be performed in such a manner that the stated LL0s will be achieved under routine conditions with typical values of 6, V, Y, and A t for the radionuclides Nn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. Oc'casionally background fluctuations, unavoidably small sample sizes, interfering radionuclides, or other uncontrollable circumstances may render these LLD's unacnievable.

j

When calculating the LLD for a radionuclide determined by gamma ray

.. . spectrometry, the backgrouno may include the typical contributions of other radionuclides normally present in the samples. The background count rate of a Ge(Li) detector is determined from background counts that are

' determined to be within the full width of the specific energy band used for the quantitative analysis for that radionuclide,

b. The principal gamma emitters for which the LLO specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and hn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that cnly these nuclices are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Material Release Report. Radionuclides which are below the LLD for the analysis should not be reported.

c. The noble gas radionuclides in gaseous effluent may be identified by either on-line (gamma spectrum) analysis of a flowing sample of effluent or by taking a grab sample of effluent and analyzing it.

d. In the event the reactor power level increases more than 15 percent in one hour and the Stack noble gas radioactivity monitor shows an activity increase of more than a factor of three after factoring about the effect due to the change in reactor power, an on-line analysis for noble gas i radionuclides in Stack effluent shall be performed on a grab sample of Stack effluent shall be collected and analyzed.

e. A composite particulate sample shall include an equal fraction of at least one particulate sample collected during each week of the compositing period.

f. In'the event a sample is collected for 24 nours or less, the LLO may be increasea by a factor of 10.

4 f

Basis: 4.6.A- Th2 purpose of sampling and analyzing reactor coolant is to.get an indication of' fuel integrity and to detennine whether Specification 3.6.A is exceeded. The radioiooine concentration in reactor coolant is' expected to change only gradually over several days. The trend of the main condenser offgas radioact,1vity is also an indicator of the trend of radioiodine in the re~ actor coolant.

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f. ,' .. 4.14 Solid Radioactive Waste Applicability: During processing wet radioactive wastes destined for disposal by-land burial as class B or C waste.

Objective: To verify that class B or C radioactive solid waste

- satisfies stability requirements before disposal.

Specification: A. Assessment to verify that class B or C radioactive waste. satisfies stability requirements in 10 CFR Part 61.56(b) before.

delivery to a carrier for transport to a licensed disposal facility shall be performed according.to the Process-Control Plan.

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t 4.15 Radioactive Effluent Monitoring Instrumentation Applicability: States surveillance requirements for OPERABILITY of radioactive effluent monitoring instrumentation.

Objective: .To demonstrate the OPERABILITY of radioactive effluent monitoring instrumentation.

Specification: A. Liquid Effluent Instrumentation Each radioactive liquid effluent monitoring instrument channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.15.1.

8. Gasenus Effluent Instrumentation Each radioactive gaseous effluent monitoring instrument channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.15.2.

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TABLE 4.15.1 .-

RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Ce heck Ca on Re a

1. Gross Radioactivity Monitors D I

a. Liquid Radwaste Effluent Line D D' R' Q b

b. Reactor Building Service Water '

E g SystemEffluentLine D H R' Q

c. Turbine Building Sump No. 1-5 0 M R' @ b

2. Flow Rate Heasurement Devices H

a. Liquid Radwaste Effluent Line D N.A. R Q c i

I TABLE 4.15.1 NOTATIONS

2. The CHANNEL FUNCTIONAL TEST shall also demonstrate that Control Room Alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm setpoint.

2. Instrument indicates a downscale failure.

3. Instrument controls not set in operate mode.

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, .. TABLE 4.15.1 NOTATIONS

a. Instrumentation shall be OPERABLE and in service excent that a channel may be taken out of service for.the purpose of a check, calibration, test or maintenance without declaring it to be' inoperable.

b. During releases via this pathway.

c. .At all times.-

d. The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and alarm annunciation in the Radwaste Control Room occur if the instrument indicates measured levels above the trip setpoint.

e. The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm setpoint.

2. Instrument indicates a downscale failure.

'3. Instrument controls not set in operate mode.

f. .The CHANNEL CALIBRATION shall be performed according tc astablished station calibration procedures.

g. On any day during which a release is made, a SOURCE CHECK shall be made at least once, before the first release.

- h'. A CHANNEL CHECK shall consist of verifying indication of flow during effluent release. A CHANNEL CHECK shall be made at least once during any day on which a release is made.

i. The CHANNEL FUNCTIONAL TEST shall also demonstrate that Control Room Alarm annunciator occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm setpoint.

2. Instrument indicates a downscale failure.

(

TABLE 4.15.2 ,..

RADI0 ACTIVE GASEOUS EFFLUENT MCNITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Channel Instrument Channel Source Channel Functional Surveillance

• Check Check Calibration Test Ecoutred

1. Main Condenser Offgas Treatment System Hydrogen Monitor D N.A. Q" M c

2. Main Stack Monitoring System (RAGEMS)

a. Radioactive Noble Gas Moniter D M R' Q' b

b. Iodine Sampler W N.A. N.A. N.A. b

c. Particulate Sampler W N.A. N.A. N.A. b

! d. Effluent Flow Measuring Device D N.A. R Q b

e. Sampler Flow Measuring Devfee D N.A. R Q b

3. Turbine Building Ventilation Monitoring System (RAGEMS)

a. Radioactive Noble Gas Monitor D M R' Q' b

b. Iodine Sampler W N.A. N.A. N.A. b

c. Particulate Sampler D N.A. N.A. N.A. b

d. Effluent Flow Measuring Device D N.A. R Q b

e. Sampler Flow Measuring Device D. N.A. R Q b G

~

. TABLE 4.15.2 NOTATIONS 4

~

J a.-; Instrumentation shall be OPERABLE and'in service except that a channel may be taken out'of service for_the purpose of,a check, calibration, test, or maintenance,,or sample media change without declaring it to be inoperable.

- b.' .During releases via this pathway.~

-l, c.-.During main condenser offgas treatment system operation.'

~

' d. During operation 'of theLeondenser air ajector. l

e. The. CHANNEL FUNCTIONAL-TEST shall also demonstrate that control room alarm annunciation occurs if any.of the following conditions exists:- ,

~1. Instrument indicates measured levels above the. alarm setpoint. i

12. Instrument indicates a downscale failure.

'3. Instrument controls not set.in operate mode.

f. The CHANNELECALIBRATION shall be performed according to established station

. calibration procedures.

g. The CHANNEL l CALIBRATION shall, include the use of at least two standard gas samples,'each'containing a known volume percent hydrogen in the range of the instrument, Dalance. nitrogen..

O 1

i

'^

4 TABLE 4.15.2 MINIMUM TEST FREQUENCIES FOR TRIP SYSTEMS Trip System Minimum Test Frequency

1) Dual Channel (Scram) Same as for respective instru-mentation in Table 4.1.1

2) Rod Block " "

3) Containment Spray, 1/3 mo. and each refueling each trip system, one at a time outage

4) Automatic Depressur12ation, Each refueling outage each trip system, one at a time

5) MSIV Closure, each closure logic Each refueling outage circuit independently (1 valve at a time)

6) Core Spray, each trip system, one 1/3 mo. and each refueling at a time. outage

7) Primary Containment Isolation, each Each refueling outage closure circuit independently (1 valve at a time)

8) Refueling Interlocks Prior to each refueling operation

9) Isolation Condenser Actuation Each refueling outage and Isolation, each trip circuit independently.(1 valve at a time)

10) Reactor Building Isolation and Same as for respective i

SGTS Initiation instrumentation in Table 4.1.1

11) Condenser Vacuum Pump Isolation Prior to each startup

12) Air Ejector Offgas Line Each refueling outage b

4.16 RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE Applicability: Environmental surveillance of radiation and radioactive effluent form the OCNGS.

Objective: Measurement and assessment of radiation and radioactive material in the environment which was discharged from the OCNGS.

Specifications: 4.16.A Radiological Environmental Monitoring Environmental samples shall.be collected and analyzed according to Table 4.16.1 Analytical techniques shall be used such that the aetection capabilities indicated in Table 4.16.2 are achieved.

Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of the automatic sampling equipment and other legitimate reasons beyond the control of GPUN. If specimens are unobtainable due to sanpling equipment malfunction, every effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from th'e sampling schedule shall be documented in the annual report.

Any location from which environmental samples or dosimetry can no longer be obtained may be dropped from the surveillance program upon notifying the NRC in writing, in lieu of any other report, that they are no longer obtainable at that location.

\

, GPU shall establish a replacement sanpling or dosimetry location 'and shall revise the ODCM in accordance with Specification 6.18.

3 If a confirmed measured radionuclide concentration in an environmental l sanpling medium averaged over any quarter sanpling period exceeds the reporting level given in Table 4.16.3, a written report shall be submitted to the NRC within sixty days of the end of the quarter during which the licensee received confirmation that a radiological limit was exceeded. If it can be demonstrated that the level is not a result of plant effluents (i.e., by conparisons with control station, natural radioactivity, or pre-operational data) a report need not be submitted, but an explanation shall be given in the Annual Radiological Environmental Report. When more than one of the radionuclides in Table 3 are detected in the medium, the reporting level shall have been exceeded if:

concentration (1) + conceItration (2) +.. 21 reporting level (1) reporting level (2)

If radionuclides other than those in Table 3 are detected and are due to i plant effluents, a reporting level is exceeded if the potential annual dose to an individual is equal to or greater than the design objectives of 10 CFR 50, Appendix I. ' Itis report shall include an evaluation of any release conditions, environmental factors, or other aspects necessary to explain the anomalous result.

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. 4.16.B .Interlaboratory Comparison Program me laboratories of the licensee and licensee's contractors which analyze radiological environmental samples shall participate in the Environmental Protection Agency's (EPA's) environmental radioactivity intercomparisons studies (crosscheck) program or an equivalent NRC-approved program.

3 A confirmatory re-analysis of the original, a duplicate, or a new sample may be desirable as appropriate. S e results of the confirmatory analyses should be completed at the earliest time consistent with the analysis, but in any case within sixty days.

- --- - - - - - - - *g y_ ,-- - -y.- .- - y.- -, - ---- wr=

4.16.C. Iand Use Survey A land use survey shall be conducted annually during the growing season to determine the location of the nearest milk. animal and nearest garden greater than 50

' square meters.(500 square feet) producing broadleaf vegetation in each of the sixteen meteorological L sectors within a distance of 8, kilometers (5 miles),1 and the locations of all milk animals and gardens greater than 50 square meters producing broadleaf I

vegetation out to a distance of 5' kilometers (3 miles) for each radial sector. If it is learned from this survey that the milk animcls or gardens are present at a location which yields a calculated thyroid dose at  ;

least 20 percent greater2than those previously sampled, or if the survey results in changes in the _

i location used in the radioactive effluent technical specifications for dose calculations, a written report ,

e'Gil be submitted to the Director of Operating Reactors, NRR (with a copy to the Director of the NRC Regional Office) within 60 days identifying the new location (distance and direction). Milk animal or-garden locations resulting in at least 20 percent higher calculated doses may then be dropped from the surveillance program. If the survey reveals that milk animals are not present or are unavailable for sanpling, then broadleaf vegetation shall be sampled.

(1). Broadleaf_ vegetation sampling may.be performed near the site boundary in a sector with the highest D/Q in_ lieu of the garden census.

(2)' -As calculated according to the ODCM.

L-

TABLE 4.16.1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Minimum Number of Sampling and Medium Sampled Sampling Locations Collection Frequency Analysis Type Airborne 1) Particulate 4 Indicator /l Background Biweekly Gross Beta Quarterly Gamma Isotopic

2) Iodine 4 Indicator /l Background Weekly I-131 Ganma radiation 30 Indicator /3 Background Quarterly Gamma Dose (TLD)

Groundwater 1 well with 2 well Quarterly Isotopic Ganna points, indicator and Quarterly H-3

Background

Surface Water 1 Indicator /l Background Monthly Gamma Isotopic Sediment 1 Indicator /l Background Semi-Annually r.asmaa Isotopic Fish 1 Indicator /l Background Semi-Annually Gamma Isotopic (when available)

Shellftsh (Clams) 1 Indicator /l Background Semi-Annually Ganna Isotopic (when available)

Food Products / Ingestion 1 Indicator (where available) Monthly Ganna Isotopic I Background (where available) (when available)

. TABLE 4.16.1 NOTATIONS

a. The LLO is defined, for purposes 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 percent probability with onlly 5 percent probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation:

LLD = 4.66 sb E . V . 2.22 . Y , exp(-Aat)

Where.

LLO is the lower. limit of detection (picocuries per unit mass or volume),.

sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute),

E is the counting efficiency (counts per disintegration),

2.2 is the number of disintegrations per minute per curie, y is the fractional radiochemical yield, when applicable, A is the radioactive decay constant for the particular radionuclide, and at for' environmental samples is the elapsed time between sample collection, or end of the sample collection period, and time of counting Analyses shall be performed in such a manner that tne stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be iaentified and described in tne Annual Radiological Environmental Report pursuant to Specification 6.9.1.e.

b. For a sample of drinking water

c. For a sample of water not used as a source of drinking water.

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L a w 8 I

B S N

L o rm i/

2 2

2 A U A1 0 0 0 P Cp 1 1 1 A . x x x C 2 ( 5 6 7 N e O l I b T a )

C T r E e T rt E  : ei D e tl c a/ *C C r Wi 00 5 0 5 5 0 0 5 5 8 5 0 0 u C 00 1 3 1 1 3 3 1 1 1 1 6 0 o p 00 0 S ( 23 1

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ep 4 7 0 0 o 4 9 0 8 5 5 5 3 3 4 4 1 t 5 5 6 5 6 9 9 1 1 1 1 3 o 3 - - - - - - - - - - - 1 s - n eF o o n r b s s a a -

I M M C C 2 Z N C C L B I

TABLE 4.16.3 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS ,

'(Source- Table 4. USNRC Branch Technical Position. Rev. 1. Nov. 79

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Resortina Level (RL)

Water Airborne Particulate Fish Milk Broad Leaf

' Analysis (pct / liter) or Gases (pct /m*) pCi/Kg wet) (pct /1) Vegetation foC1/Ka. wet!

H-3 2 x 10*(a)

Mn-54 1 x 10 3 3 x 10*

Fe-59 4 x 10 2 1 x 10*

Co-58 1 x 10 8 3 x 10*

Co-60 3 x 102 1 x 10' Zn-65 3 x 102 2 x 10 4 Zr-Nb-95 4 x 10 2 I-131 2 0.9 3 1 x 10 2 Cs-134 30 10 1 x 10 3 60 1 x 10 5 Cs-137 50 20 2 x 10 8 70 2 x 10 3 Ba-La-140 2 x 10 2 3 x 10 8

a. Well Water Only

., BASES Basis: 4.16.A It should be noted that in addition to the sampling and analysis required by_the proposed technical specifications, GPU Uuclear may choose, to conduct additional sampling and analysis as deemed advisable to assure adequate protection of the health and safety of the public and monitoring of the environment. The

" Pathway to Man" concept is emphasized throughout, and the resultant program is directed toward evaluating those media, locations, isotopes, etc. that affect the radiological impact on man. The specified detection capabilities are presently state-of-the-art for routine environmental measurements in industrial laboratories and match those suggested in the USNRC Branch Technical Position on Radiological Environmental Monitoring.

Basis: 4.16 C GPUN may propose any of the following methods to accomplish the land use survey. These methods are generally listed in order of overall preference - considering quality of data, cost, and speed of accomplishment on an annual basis. Interpretation of aerial photographs may be the most desirable method for accomplishing an annual land use survey within the vicinity of the Oyster Creek Nuclear Generating Station. In addition to this, information from local and state government agencies will be utilized. Door to door census in the vicinity of Oyster Creek Nuclear Generating Station are not usually a desirable way to produce land use infonnation because of the high number of seasonal / rented residencies in this area. In addition, the high number of dwellings would require an inordinate manpower effort to accomplish a complete census on an annual basis. GPUN may elect, however, to conduct field checks of selected areas that are not fully understood after the interpretation of aerial photographs and the use of state and local government data.

p la. Tha ccnformance of facility operation to all provisions contain:d within tha Technical Specifications and applicable licer.se conditions at least once per year.

b. ~The performance training'and qualifications of the entire facility staff at least once.per year.

c. LThe-results of all actions taken to correct-deficiences occurring in facility equipment, structures, systems or method of operation that affect. nuclear safety at least once per six months.

d. The Facility Emergency. Plan an'd implementing procedures at least once perstwo years.

e. The Facility. Security Plan and implementing procedures at least

.once per two years.:

f.- The radiological environmental monitoring program and the results thereof at least once per 12 months.

~

g. The 0FFSITE 00SE' CALCULATION MANUAL and implementing procedures at

'least once per 24 months.

h.- LThe' PROCESS CONTROL PROGRAM and implementing procedures for radioactive wastes at.least once per 24 months.

i.. Any other area of facility operation considered appropriate by the

. :GORB or the Vice President and Director.

6.5.3.6 RECORDS' Written documentation of all independent safety reviews and Investigations will be forwarded.to the Vice President and Director and the Chairman of the General Office Review Board. In addition, any reportable occurrence or item involving an unreviewed safety question which is identified by the ISRG will be documented and reported immediately to the above mentioned persons.

'The audit findings which result from all audits conducted in accordance with Section 6.5.3.5 shall be documented and reported to the above mentioned persons within 30 days after completing the audit. Reports documenting corrective action will receive the same distribution and they will also be forwarded to the ISRG Coordinator.

6.5.4 GENERAL OFFICE REVIEW 8OARD (GURB) 6.5.4.1 FUNCTION-The technical and administrative function of the G0RB is provide

' independent review of major safety issues, to foresee potentially significant nuclear'and radiation safety problems, and to advise the Office of the President on these matters.

6.5.4.2 COMPOSITION Members of the General Office Review Board shall possess extensive experience in their incividual specialities and collectively have the competence in the following areas:

a. Nuclear Power Plant Operations

b. Nuclear Engineering

c. Chemistry and Radiochemistry

d. Metallurgy-

e. Instrumentation and Control

f. Radiological Safety

g. Mechanical and Electrical Engineering The Chairman and Vice Chairman shall be appointed by the Office of the President. (Neither shall be an individual with line responsibility for operation of the plant).

The Chairman shall designate a minimum of six additional members.

Not more than a minority of the Board shall have line responsibility for operation of Oyster Creek Nuclear Generating Station.

6.5.4.3 ALTERNATES Alternate members shall be appointed in writing by the GORB

-Chairman and will have the type'of experience and training required

.of regular members, however, they need not have the extensive longevity in the designated fields as long as, in the opinion of the Chairman, their experience and judgement are adequate.

6.5.4.4 MEETING FREQUENCY The GURB shall meet at least semi-annually and any time at the request of the Chairman or the Office of the President.

6.5.4.5 QUORUM A quorum shall consist of the Chairman or Vice Chairman and three members / alternates. No more than one alternate member shall be counted when establishing a quorum and no more than a minority of the quorum shall hold line responsibility for operations of the Oyster Creek Station.

6.5.4.6 RESPONSIBILITIES

a. The primary responsioility of the GOR 8 is to foresee potentially significant nuclear and radiation safety problems and to recommend to the Office of the President how they may be avoided or mitigated.

b. Carry out the specific independent safety review responsibilities listed in Table 6.5-1.

J

, i. 6.5.4.7 AUTHORITY

. The G0RB shall be advisory to the Office of the President and shall have the authority to conduct reviews, audits, and investigations requested by the Office of the Presiaent or as deemed necessary by-the GORB in the fulfillment of its responsibilities.

6.5.4.8 AUDITS The report of the management review of the QA Plan, initiated by the Vice President and Uirector in accordance with the Operational Quality Assurance Plan, shall be reviewed by the G0RB with respect to safety and administrative safety issues.

6.5.4.9 RECORDS Minutes of each GORB meeting shall be recorded and approved by the GOR 8 Chairman. Copies of approved minutes will be forwarded to the Office of the President, Vice President and Director, PORC Chairman, and others designated by the G0RB Chairman. GORd recommendations to the Office of the President will be documented in a letter from the GORB Chairman to the Office of the President.

Included with each letter will be.any dissenting opinions of members of the Board.

6.6 REPORTABLE OCCURRENCE ACTION 6.6.1 The following actions shall be taken in the event of a Reportable Occurrence:

a. The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.

b. Each Reportable Occurrence Report submitted to the Commission shall be reviewed by the Plant Operations Review Committee and submitted to the ISRG Coorcinator and the Vice President and Director.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in tne event a Safety Limit is violated:

a. If any Safety Limit is exceeded, the reactor shall be shut down immediately until the Commission authorizes the resumption of operation.

b. The Safety Limit violation shall be reported to the Commission and the Vice Presiaent and Director.

Print LM/003 Page 1 of 2 TA8tE 6.5-1 -

SAFETY REVIEW RESPONSISILITIES INDEPENDENT REVIEW ITEM INITIAL ACITON PORC ISRG race 4

a. Pr: posed change to equipment, or Initiator: Must prepare a complete Must review items to Hust review all May review any deter-system subject to the Provisions of description of the proposed changes determine whether or determinations by mination, but must S;ction 50.59. Part 50. Title 10 and ensure a safety evaluation of the not an unreviewed the Vice President review those for which Code of Federal Regulations. change is included. safety question is and Director, the Vice President and Vice President & Director (1) Must involved. If Director has requested determine if the item is an actual requested by the GORB review.

change to equipment or systems as Vice President described in the FSAR. (2) Must and Director.

determine if the item involves an unreviewed safety question. (3) May request the PORC to assist in the above determinations.

b. Prcposed tests and experiments. Initiator: Must prepare a complete Must review item to As above. As above.

(subject to provisions of) 50.59, description of the proposed test or determine whether or Part 50. Title 10. Code of Federal experiment and ensure a safety not an unreviewed R:gulations. evaluation of the test or experiment safety question is is included. involved, if Vice President & Director (1) Must requested by the determine if the item involves an Vice President unreviewed safety question. (2) May and Director.

request the PORC to assist in the above determinations.

c. Preposed changes in Technical Initiator: Must prepare a complete (1) Must review the Must review change May review any item but Sptcifications or in tne NRC description of the proposed change item for nuclear and and PORC reconnend- must review those for Optrating License. and ensure a safety evaluation of the radiological safety. atton prior to which the change is included. (2) Must make submittal to the NRC. Vice President and recommendations to Director or his l the Vice President supervisor have and Director as to requested GOR 8 review.

whether or not the change is safe.

d. Reportable occurrences Vice President and Director: Must have Must review Report- Must review Report- As above.

l' investigations performed for all able Occurrences able Occurrence Reportable Occurrences and a report Report for safety Reports for safety prepared including the safety significance and significance and signifiance of the incident. make reconnendations review PORC to the Vice reconnendations.

President and Director on how to avoid recurrence.

1

+ .

Page 2 of 2 TABLE 6.5-1 SAFETY REVIEW RESPONSI8ILITIES  ;

^

INDEPENDENT REVIEW INITIAL ACTION PORC ISRG r_nna ITEM ,

e. Fact 11ty operations including Continuing respond- See item i below. As above.

Security Plan. Emergency Plan and ability.

Implementing procedures; review is to detect potential safety hazards.

f. Significance operation abnormalities Vice President and Director report Review matter and Perform independent As above, er deviations from normal and such matters to the PORC ISRG report evaluation review of PORC axpected performance. Coordinator and the Chairman GOR 8. of safety evaluation.

significance to the ISRG and GGRS.

As above. As above. As above,

g. Any indication of an unanticipated As above.

dificiency in some aspect of design or operation of safety related structures, systems or components.

h. PORC minutes and reports. Review to determine As above.

if any matters discussed involve unrelated safety questions.

1. Audit Reports and NRC Inspection Review to determine The report of the Reports. If any matters management review of reported involve the QA Plan, Initialed Violations of by the Vice President Technical Specifica- ar.d Director. Oyster tions License Creek in accordance requirements or have with the Operational any nuclear or Quality Assurance Plan radiation safety shall be reviewed by impitcations. the GOR 8 with respect to technical and administrative safety Issues.

j. Prccess Control Plan Review changes, advise Vice President and Director.

k. Offsite Dose Calculation Manual Review changes, advise Vice President and Director.

Amendment No. 65

, . c. A Safcty Limit Violaticn Rep;rt shall be prepar:d. The report shall

. be reviewed by the Plant Ooerations Review committee and submitted to the Vice President and Director. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components systems or structures, and (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Commission within 10 days of the violation. It shall also be submitted to the ISRG Coordinator.

6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented, and maintained that meet or exceed the requirements of Section 5.1 and 5.3 of American National Standard N18.7-1972 and Appendix "A" of the Nuclear Regulatory Commission's Regulatory Guide 1.33-1972 except as provided in 6.8.2 and G.8.3 below.

Written procedures shall be adopted and maintained to implement the:

Process Control Plan I

Offsite Dose Calculation Manual 6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the Plant Operations Review Committee

, and approved by the Vice President and Director prior to implementation and periodically as specified in the Administrative Procedures.

6.8.3 Temporary changes to procedures 6.8.1 above may be made provided:

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

b. The change is approved by two members of the supervisory staff, at least one of whom possesses a Senior Reactor Operator's License.

c. The change is documented, subsequently reviewed by the Plant Operations Review Committee and approved by the Vice President and Director as specified in the Administrative Procedures.

6.9 REPORTING REQUIREMENTS

, 6.9.1 a. Startup Report: A summary report of plant startup and power

escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a plannea increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel sucolier, and (4) modifications that may have significantly altereo the nuclear, thermal, or hydraulic performance of the plant. The report shall address each of the tests identified in the FSAd and shall in general include a description of the measured values of the operating conaitions; or characteristics

obtained during the test program and_a comparison of these values

. with design pr: dictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also De described. Any additional specified details required in license conditions based on other commitments shall be included in this report.

Startup reports shall be submitted within (1) 90 days followir.g completion resumption orofcommencement the_startup test program, (2) of commercial 90 operation, power days following(3) or 9 months following initial criticality, whichever is earliest.

If the Startup Report does not cover all three events (i.e.,

initial criticality, completion of startup. test program, and resumption or commencement of commercial power operation),

supplementary reports shall be sutimitted atileast every three months until all three events have been completed.

Prior to startup of each cycle, a special report presenting the results of the in-service inspection of the Core Spray Spargers during each refueling outage shall be submitted to the Commission for review.

~

b. Annual Exposure Data Report: Routine exposure data reports

. covering the operation of the unit during the previ,ous calendar 3 year shall be submitted prior to March 1 of each year. Reports) shall contain a tabulation on~an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem / year and their associatea man rem exposuie according to work and job functions (This tabulation supplements the regoirements of 10' CFR 20.407),

e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), '

waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less tnan 4 20 percent of the individual total dose need not be accounted

.for. In tne aggregate, at least 80 percent of the total whole body dose received from external sources shall be assigned to specific major work functions.

c. Monthly Operating Reoort: Routine reports of operating statistics and shutdown experience shall De submitted on a monthly basis which will include a narrative of operating experience, to the Director, Office of Management and Program Control, U.S. Nuclear Regulatory Commission, Washington, D.C.

20555, with a copy to the Regional Office of Iad, no later than the 15the. of each month following tha calendar month covered by the report. .

- - - - , , - - - - - . , ~ g --

k Semiannual Radioactive Material Release Report: A report of

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i -

radioacilve materials releasea trom the Station during the

[

preceedi.;g six months shall be submitted to the NRC within 60

' days after January 1 and July 1 of each year. Each report shall include the following information:

1. a summary by calendar quarter and by radionuclide of the

. quantities of radioactive liquid and gaseous effluents from the Station,

'2. a summary of radioactive solia waste shipped from the Station including:

a. physical description of the waste

b. classification of the waste, per 10 CFR Part 61

c. solidification agent

.d. total volume shipped

e. total quantity of radioactive material shipped (curies)

'f. identity of principal radionclides,

3. a description of any changes to the PCP or ODCM,

4. a summary of meteorological data collected during the year shall be included in the report submitted within 60 aays after January 1 of each year.

e. Annual Radiological Environmental Report: A repcrt of radiological environmental surveillance activities during each year.shall be submitted before May 1 of the following year. Each report shall include the following information required-in Specification 4.16 for radiological environmental surveillance:

2

1. a summary description of the radiological environmental monitoring procram,

~

2. a map and a table of distances and directions (compass azimuth) of locations of sampling stations from the reactor,

3. resu,lts of analyses of samples and of radiation measurements, (In tne event some results are not available, the reasons shall be explained in the report. In the event the missing results are stained, they shall be reported to the MC as soon as is reasonable.)

4. deviation (s) from the environmental sampling schedule in Table 4.16.1.

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5. a report of analyses which did not achieve the LLO required in Table 4.16.2.

b,

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6. a summary of the results of the land use survey,

7. a summary of the results of licensee participation in an NRC approved inter-laboratory crosscheck program for environmental samples if not participating in an EPA crosscheck program,

8. results of dose evaluations to assess compliance with 40 CFR Part 190.10a.

Basis: 6.9.1.e An annual report of radiological environmental surveillance activities includes factual data summarizing results of activities required by the surveillance program. In order to aid interpretation of the data, GPUN may choose to submit analysis of trends and comparative non regional radiological environmental data. In addition, the licensee may choose to discuss previous radiological environmental data as well as the observed radiological environmental impacts (f station operation (if any) on the environment.

In the event OCNGS or its contracted laboratory for environmental analyses participates in the EPA's interlaboratory comparison program, it is the intention of GPU Nuclear to pursue the option offered by the NRC in Rev. I to the Branch Technical Position dated November, 1979 i.e.,

to provide their EPA program code so that the hRC can review the EPA's participant data directly in lieu of actually submitting the data in the annual report. Otherwise, the results will be summarized in the annual report according to 6.9.1.e.7.

6.9.2 REPORTABLE OCCURRENCES Reportable occurrences, including corrective actions and measures to prevent reoccurrences, shall be reported to the NRC. Supplemental reports may be required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a license event report shall be completed and reference shall be made to the original report date.

a. Proiapt Notification with Written Followup. The types of events listed below shall be reported as expeditiously as possible, but within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegraph, mailgram, or facsimile transmission to the Director of the appropriate Regional Office, or his designate no iater than the first working day following the event, with a written followup report within two weeks. The written followup report shall include, as a minimum, a completed copy of a licensee event report fonn. Information provioed on the licensee event report

% form sh'll a bn. supplemented, as ne:ded, by' additional narrative

. _ material to provioe a complete explanation of_the circumstances i surrounding the' event.,

1. Failure _of'the reactor-protection system or'other systems subjectcto; limiting safety system' settings to initiate the required _ protective function by the time a monitored parameter reaches the setpoint specified as the limiting

- system setting in-the technical specifications or failure to

- complete the required protective function. 1

. NOTE:~ Instrument drift discovered as a result of testing need not be reported undeer this item but may be reportable F under items 2.a(5), 2.a(6), or 2.o(1) below.

. 2. Operation of the unit or.affected systems when any parameter or operation subject to a_ limiting condition is less conservative than the least conservative aspect of the limiting condition for operation _estaolished in the technical

. specifications..

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-NOTE: If specified action is taken when a system is found to

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be operating'between the most conservative and the

'least conservative aspects of-a limiting condition for p

operation listed in the technical specifications, the limiting condition for operation is not considered to' have been violated and need not be reported under this item, but it.may_be~ reportable under item 2.b(2) below.

, 3. Abnormal. degradation discovered in fuel cladding, reactor i coolant pressure boundary, or primary containment.-

NOTE: Leakage of valve packing or gaskets within the limits for identifiea leakage set forth in technical specifications need not be reported under this item.

4. Reactivity.abnormalties, involving disagreement with the predicted value of reactivity balance under steady state conditions during power operation, greater than or equal to-1 percent AK/k; a calculated reactivity balance. indicating shutdown margin less conservative than specified in the

. technicallspecifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds-4 or, if sub-critical, an unplanned reactivity' insertion of more than 0.5 percent ak/k or occurrence of any unplanned criticality.

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5. Failure or malfunction of one or more components which

, prevents or could prevent, by itself, the fulfillment of the functional requirements of system (s) used to cope with accidents analyzed in the SAR.

6. ' Personnel error or procedural inadequacy which prevents or

., could-prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the SAR.

NOTE: For items 2.a(5) and 2.a(6) reouced redundancy that does not result in a loss of system function need not be reported under this section but may be reportable under items 2.b(2) and 2.b(3) below.

7. 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.

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

9. Performance of structures, systems, 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.

NOTE: This item is intended to provide for the reporting of potentially generic problems.

b. Thirty Day Written Reports. The reportable occurrences discussed below shall be the subject of written reports to the Director of the appropriate Regional Office within thirty days of occurrence of the event. The written report shall include, as a minimum, a complete copy of a licensee event report form. Information provided on the licensee event form shall be supplemented, as needed, by additional narrative material to provice-complete explanation of the circumstances surrounding the event.

1. Reactor orotection system or engineered safety feature instrument settings which are found to be less conservative than those established Dy the technical specifications but which do not prevent the fulfillment of the functional requirements of affected systems.

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2. ~C:nditions lea' ding to operation in a degrad:d mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation.

NOTE: Routine surveillance testing, instrument calibration, or preventive maintenance which require system configurations as described in items 2.b(1) and 2.b(2) need not be reported except where test results themselves reveal a degraded mode as described above.

3. Observed inadequacies in the implementation of administrative

, or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems.

4. Abnormal degradation of systems other than those specified in Item 2.a(3) above designed to contein raoicactive material resulting from the fission process.

NOTE: Sealed sources or calibration sources are not included under this item. Leakage of valve packing or gaskets within the limits for identified leakage set forth in technical specifications _need not be reported under this item.

6.9.3 UNIQUE REPORTING REQUIREMENTS Special reports shall be submitted to the Director of Regulatory Operations, Regional Office within-the time period specified for

.each report. These reports shall De submitted covering the activities identified below pursuant to the requirements of the applicable reference specification.

a. Materials Radiation Surveillance Specimen Reports (4.3A)

b. Integrated Primary Containment Leakage Tests (4.b)

c. Results of required leak tests performed on sealed sources if the tests reveal the presence of 0.005 microcuries or more of removeable contamination.

d. Inoperable Fire Protection Equipment (3.12)

e. Core Spray Sparger Inservice Inspection (Table 4.3.1-9)

Prior to startup of each cycle, a special report presenting the results of the inservice inspection of the Core Spray Spargers during each refueling outage shall be submitted to the Commission for review,

f. Liquid radwaste batch discharge exceeding Specification 3.6.B.1.

g. Main condenser offgas discharge without treatment per Specification 3.6.0.1.

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' h. Dose dun to radiot'ctive liquid effluent exceeding.

Specification-3.6.J.l.

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1. Air dose dua to radioactive noble gas in gaseous effluent-

' exceeding Specification'3.6.L'.l.

. j.: Air dose due to radioiodine and particulates exceeding-

~ Specification 3.6.M.l.

k.-Annual.totalJdose due to radioactive effluents exceeding

" Specification 3.6.N.l.

1. Records of.results of analyses required by the_ Radiological Environmental Monitoring Program.

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'6 .10 R'ECORD RETENTION -

T 6.10.1 The.following records shall_be retained for a least five years:

a. Records 'and logs of facility operation covering time interval at p each power level.

- b. Records and logs of principle maintenance activities,

. inspections,: repair and replacement of principal items of'

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equipment relatea to nuclear safety. '

c. Reportable occurrence reports.

d. Records of surveillance activities, inspections and calibrations required by these technical specifications.

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l ~e. Records of: reactor tests 1and experiments.

g f.. Records of changes made to operating' procedures.

. g. Records of radioactive shipments.

h. Records,of sealed-source leak tests ~and results.

i. Records of annual physical. inventory of all source material of record.-

- 6.10.2 The-following records shall be retained for the duration of the

- Facility Operating License:

a. Record and drawing changes reflecting facility design _

modificatio'ns made to systems and equipment described in the. "

Final Safety Analysis Report.

<_ b. Records of new and~ irradiated fuel inventory, fuel transfers and

_ assembly burnup histories. >

c.' Records of facility radiation and contamination surveys.

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d. Records of radiation exposure for all individuals entering

. radiation control areas.

e. Records of gaseous and liquidLradioactive material released to the environs.

f. Records of transient or operational cycles for those facility components designed for a limited number of transients or cycles.

g. Records of training and qualification for current members of the plant staff.

h. Records of inservice inspections perfonned pursuant to these technical specifications.

i. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

J.-Records of meetings of the Plant Operations Review Committee and

'the General Office Review Board.

k. Records for Environmental Qualification which are covered under the provisions of paragraph 6.14.

1. Records of results of analyses required by the Radiological Environmental Monitoring Program.

6.10.3 Quality Assurance Records shall be catained as specified by the Quality Assurance Plan.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

6.12 DELETED- ,

6.13 HIGH RADIATION AREA 6.13.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously. posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP).

NOTE: Health Physics personnel shall be exempt fonn the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they are following plant radiation protection procedures for entry into high radiation areas.

An individual or group of individuals permitted to enter such areas shall be provided with one or more of the following:

a. A radiation monitoring device which continously indicates the radiation dose rate in the area.

b. A radiation monitoring device which continuously intergrates the radiation dose rate in the area and alarms when a pre-set integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them,

c. A health physics qualified individual (i.e. qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive exposure control over the activities within the area and who will perform periodic radiation surveillance at the frequency in the RWP. The surveillance frequency will be established by the Radiological Controls Manager.

6.13.2 Specification 6.13.1 shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of operations and/or radiation protection supervision on duty.

6.14 ENVIRONMENTAL QUALIFICATION A. By no later than June 30, 1982 all safety-related electrical equipment

D in the facility shall be qualified in accordance with the provisions of: Division of Operating Reactors " Guidelines for Evaluating Environmental Qualification of Class 1 E Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588 " Interim Staff

. Po'sition on Environmental Qualification of Safety-Related Electrical Equipment", Decemoer 1979. Copies of these documents are attached to Order for Modification of License DPR-16 dated October 24, 1980.

8. By no later than December 1, 1980, complete and audibile records must be

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3 available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the 00R Guidelines or kUREG-0588. Thereafter, such

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records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.

6.15 INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This y program shall include the following:

'. l. Provisions establishing preventive maintenance and periodic visual inspection requirements, and c

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2. System leak test rcquirements, to.the. extent permitted by system design'

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.. . and radiological conditions, for each system at a frequency not to

' exceed refueling cycle intervals. . The systems subject to this testing are (1) Core Spray, (2) Containment Spray, (3) Reactor Water Cleanup, (4) Isolation condenser and (5) Shutdown Cooling.

6.16 I0 DINE MONITORING-The licensee shall implement a program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas

• under accident conditions. This program shall-include the following:

1. Training of personnel,

2. Procedures for monitoring, and .

3. Provisions for maintenance of sampling and analysis equipment.

Areas requiring personnel access for establishing hot shutdown condition.

6.17 PROCESS-CONTROL PLAN

.l. GPU may change the Process control Plan provided each change is submitted to the Commission by inclusion in the Semiannual Radioactive Material Release Report for the period in which the change is made effective and contains:

a. Sufficiently detailed information to support the rationale for the change,

b. a determination that the product waste-form will conform to the requirements of 10 CFR Part 61.56, and

c. documentation of review and approval of the change by the Plant Operations Review Committee.

.2.. Change (s) shal1 become effective after review and approval oy the Plant Operations Review Committee and approval by the Vice President and Director. -

.6.18 0FFSITE DOSE CALCULATION MANUAL

1. GPU may make changes to the Offsite Dose Calculation Manual (ODCM) provided each change is submitted to the Commission in the Semiannual Radioactive Material Release-Report for the period in which the change is made effective. The submittal shall contain:

a. sufficiently detailed information to support the rationale for the change, p

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b. a d;t:rmination that thi change will not substantially reduce the s ability of dose calculations or setpoint detenninations to assess compliance with Specifications, and

c. documentation of review and approval of the change (s) by the Plant Operations Review Committee.

2. Change (s) shall become' effective after review and approval by the Plant Operations Review Committee and approval by the Vice President and Director.

.6.19 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS

1. Each modification to a radioactive waste treatment system which involves an unreviewed safety question:

a. Shall be performed in accordance with the provisions of 10 CFR Part 50.59, except

b. The description of the modification and a written safety evaluation which' includes the bases for the change shall be submitted as part of the annual FSAR update, and

c. Shall become effective upon review and approval by the Vice President and Director.

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