ML20101S504

Amend 109 to License DPR-49,revising Tech Specs to Incorporate Radiological Effluent Tech Specs

ML20101S504
Person / Time
Site:	Duane Arnold
Issue date:	01/14/1985
From:	Vassallo D Office of Nuclear Reactor Regulation
To:	Corn Belt Power Cooperative, Central Iowa Power Cooperative, Iowa Electric Light & Power Co
Shared Package
ML112371143	List: ML20101S504
References
DPR-49-A-109 NUDOCS 8502050446
Download: ML20101S504 (76)

Text

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION -

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j WASHINGTON, D. C. 20555

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l IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE

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Amendment No.109 License No. OPR-49 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Iowa Electric Light & Power Company, et al, dated April 6,1983, as supplemented July 29, 1983, October 17, 1983 and July 25, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; i

and l

l E.

The issuance of this amendment is in accordance with 10 CFR Part l

51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-49 is hereby amended to read as follows:

8502050446 850114 PDR ADOCK 05000331 P

PDR

F (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 109, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

The license amendment is effective as of January 1, 1986.

FOR THE NUCLEAR REGULATORY ComISSION Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: January 14,198S

FE ATTACHMENT TO LICENSE AMENDMENT NO.109 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 Revise the Appendix A Technical Specifications as indicated below.

Remove Insert 1.0-6 through 1.0-8 1.0-6 through 1.0-9 3.2-2 through 3.2-4 3.2-2 through 3.2-4 3.?-29 3.2-29 3.2-33 3.2-33 3.2-45b 3.7-19b 3.7-48b 3.14-1 through 3.14-15 3.15-1 through 3.15-15 3.16-1 through 3.16-12 6.5-3 6.5-3 6.5-9 6.5-9 6.5-10 6.5-10 6.8-1 6.8-1 6.8-2 6.8-2 6.10-2 6.10-2 6.10-3 6.10-3 6.11-3 through 6.11-6 6.11-3 through 6.11-11 6.11-15 6.14-1 6.14-2 6.15-1 1

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l 22.

INSTRUMENTATION-a.

Instrument Calibration or Channel Calibration - An Instrument

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. Calibration means the verification or adjustment of an instrument signal output so that it corresponds, within acceptable range and accuracy, to a known value(s) of the parameter dich the instrument monitors. The acceptable range and accuracy of an instrument and its setpoint are given in the system design control document and its setpoint is used in the Technical Specifications.

Instrument calibration may be performed by any series of sequential, overlapping, or total channel steps such that the entire instrument is calibrated.

Instrument calibration includes the Instrument or Channel Functional Test, as appropriate.

b.

Channel.- A channel is an arrangement of a sensor and associated components used to evaluate plant variables and produce discrete outputs used in logic. A channel terminates and loses its identity dere individual channel outputs are combined in logic.

c.

Instrument or Channel Functional Test - An Instrument or Channel Functional Test for (1) Analog channels means the injection of a simulated signal into the channel as close to the sensor as practicable to verify the proper response, alarm, and/or initiating action.

(2) Bistable channels means the injection of a simulated signal into the sensor to verify the proper response, alarm and/or -

initiating action.

d.

Instrument or Channel Check - An instrument or channel check is a qualitative determination of acceptable operability by observation

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of instrument with other independent instruments measuring the same variable.

e.

Logic System Functional Test - A logic system functional test means a test of all relays and contacts of a logic circuit to insure all components are operable per design intent. Where practicable, action will go to completion; i.e., pumps will be l

started and valves operated.

4 f.

Trip System - A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function. A trip system may require one or more instrument channel trip -signals related to one or more plant parameters in order to initiate trip system action.

Initiation of protective action may require the tripping i

of a single trip system or the coincident tripping of two trip l

systems.

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Protection Action - An action initiated by the protection system i

when a limit is reached. A protective action can be at a channel.

j or system level.

1.0-6 Amendment No. 109 i

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22.

Instrumentation - Continued h.

Protective Function - A system protective action which results from the protective action of the channels monitoring a particular L

plant condition.

i.

Simulated Automatic Actuation - Simulated automatic actuation means applying a simulated signal to the sensor to actuate the circuit in question.

j.

Logic - A logic is an arrangement of relays, contacts, and other components that produces a decision output.

t 1)

Initiating - A logic that receives signals from channels and produces decision outputs to the actuation logic.

2) Actuation - A logic that receives signals (either from initiating logic or channels) and produces decision outputs to accomplish a protective action.

i k.

Primary Source Signal - The first signal, which by plant design, should initiate a reactor scram for the subject abnormal occurrence (see FSAR Subsection 14.5).

1.

Source Check - A Source Check is the assessment of channel response when the channel sensor is exposed to a source of radiation.

23. FUNCTIONAL TESTS t

A functional test is the manual operation or initiation of a system, i

subsystem, or component to verify that it functions within design tolerances (e.g., the manual start of-a core spray pump to verify that it runs and that it pumps the required volume of water).

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24. SHUTDOWN The reactor is in a shutdown condition when the reactor mode switch is in j

the shutdown mode position and no core alterations are being performed.

25. ENGINEERED SAFEGUARD An engineered safeguard is a safety system, the actions of which are i

essential to a safety action required in response to accidents.

26. SURVEILLANCE FREQUENCY i

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- Periodic surveillance tests, checks, calibrations and examinations shall be i

performed within the specified surveillance intervals.

These intervals may be adjusted plus or minus 25%. The operating cycle interval as pertaining to instrument and electrical surveillance shall never exceed 15 months; In cases where the elapsed interval has exceeded 100% of the.specified interval, the next surveillance interval shall commence at the end of the original specified. interval.

1.0-7 Amendment No.109

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27. FREQUENCY NOTATION NOTATION FREQUENCY 3

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 />.

D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

A At least once per year.

R At least once per 18 months.

J S/U Prior to each reactor startup.

P Prior to each release.

NA Not applicable.

28. FIRE SUPPRESSION WATER SYSTEMS A fire suppression water systen shall consist of a water source, pumps, and distribution piping with associated sectionalizing control or isolation valves. Such valves include yard hydrant curb valves, the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe or deluge systen riser.

-29.

REACTOR TRIP SYSTEM RESPONSE TIME I

Reactor trip system response time is the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until deenergization of the scram pilot valve solenoids.

30. REPORTABLE EVENT l

A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part' 50.

31. OFFSITE DOSE ASSESSMENT MANUAL l

The Offsite Dose Assessment Manual (ODAM) is a manual containing -the l

methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of radioactive, gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints.

32. GASEOUS RADWASTE TREATMENT SYSTEM A Gaseous Radwaste Treatment Systen is any systen designed'and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing delay or holdup for the purpose of reducing radioactivity prior to release to the environment.

33. PURGE - PURGING l

PURGE or PURGING is the controlled process of discharging air or gas from a i

confinement to maintain temperature, pressure, humidity, concentration or i

other operating condition, in such a manner that replacement air or gas is l

required to purify the confinement.

1.0-8 Amendment No. 109

34. VENTING VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during the process.

Vent, used in system names, does not imply a VENTING process.

35. PROCESS CONTROL PROGRAM (PCP)

The PROCESS CONTROL PROGRAM shall generally describe the essential process controls and checks used to -assure that a process for solidifying radioactive waste from a liquid system produces a product that is acceptable for burial according to 10 CFR Part 61.56.

36. MEMBER (S) 0F THE PUBLIC Member (s) of the Public are persons who are not occupationally associated with Iowa Electric Light and Power Company and who do not normally frequent the DAEC site. The category does not include contractors, contractor employees, vendors,' or persons who enter the site ta make deliveries or to service equipment.

37. SITE BOUNDARY The Site Boundary is that line beyond which the land is neither owned, nor leased, nor otherwise controlled by IELP. UFSAR Figure 1.2-1 identifies the DAEC Site Boundary. For the purpose of implementing radiological effluent technical specifications, the Unrestricted Area is that land (offsite) beyond the Site Boundary.

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1.0-9 Amendment No. 109 i

DAEC-1 LIMITING CONDITIONS FOR OPERATION-SURVEILLANCE REQUIREMENT C.

Control Rod Block Actuation C.

Control Rod Block Actuation 1.

The limiting conditions of Instrumentation shall be operation for the functionally tested, calibrated instrumentation that and checked as indicated in initiates control rod block Table 4.2-C.

are given in Table 3.2-C.

System logic shall be i

2.

The minimum-number of functionally tested as operable instrument channels

-indicated in Table 4.2-C.

specified in Table 3.2-C for the Rod Block Monitor may be reduced by one in one of the trip systems for maintenance and/or testing, provided that this condition does not last longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any thirty day period.

D.

Radiation Monitoring Systems-D.

Radiation Monitoring Systems-isoiation e initiation isoiation e initiation runct10n5 PunCtlon5 1.

Steam Air Ejector Offgas 1.

Steam Air Ejector Offgas system 3ystem a) At least one post Instrumentation shall be treatment steam air functionally tested, calibrated ejector offgas system and checked as indicated in radiation monitor shall be Table 4.2.D.

operable during reactor power operation. The System logic shall be monitors shall be set to functionally tested as initiate immediate closure indicated in Table 4.2-D.

of the charcoal delay bed 1

bypass valves if the monitor exceeds a trip setting equivalent to the dose rate specified in Specification 3.15.2.1.

b) In the event no post-i treatment monitor is operable, gases from the steam air ejector offgas system may be released to the environment for up to charcoal bed of the(1)ffgas 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided-the o

delay system is not bypassed, and (2) the offgas stack noble gas activity monitor-is operable. -

Otherwise be in at leasti HOTSTAND$Ywithinthe i

following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.2-2 AmendmentNo.L109-1

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT c) At least one pre-treatment steam air ejector offgas system radiation monitor shall be operable during reactor power operation.

The monitors shall be set to initiate an alarm if the monitor exceeds a trip setting equivalent to 1.0 Ci/sec of noble gases after 30 minutes delay in the offgas holdup line.

In the event the noble gas flow in the air ejector offgas exceeds the equivalent of 1.0 Ci/sec after 30 m.inutes delay in the offgas holdup line, restore the rate to less than this 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 standby within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

d) In the event no pre-treatment monitor is operable, gases from the steam air ejector offgas system may be released for up to 30 days provided (1) the charcoal bed of the offgas delay system is not bypassed, (2) Grab samples are collected and analyzed weekly, and (3) the offgas stack noble gas activity monitor is operable, or at least 1 post-treatment monitor is operable.

2.

Reactor Building Isolation 2.

Reactor Building Isolation and Standby Gas Treatment and Standby Gas Trea_tment system system The limiting conditions for Instrumentation shall be operation are given in functionally tested, calibrated Specification 3.7.B.

and checked as indicated in Table 4.2-D.

System logic shall be functionally tested as indicated in Table 4.2-D.

3.2 Amendment No. 109

N DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

-E.

Drywell Leak Detection E.

Drywell Leak Detection The limiting conditions of Instrumentation shall be i

operation for the calibrated and checked as instrumentation that monitors indicated in Table 4.2-E.

drywell leak detection are given in Table 3.2-E.

F.

Surveillance Information F.

Surveillance Information Readouts Readouts The limiting conditions for Instrumentation shall be the instrumentation that calibrated and checked as provides surveillance

. indicated in Table 4.2-F.

information readouts are given in Table 3.2-F.

G.

Recirculation Pump Trip G.

Recirculation Pump Trip i

(ATWS)

Instrumentation and logic shall be functionally tested, The limiting conditions for calibrated, and response time operation for the tested as indicated on Table instrumentation that trips the 4.2-3.

recirculation pumps as a means of limiting the consequences of a failure to scram during an anticipated transient are given in Table 3.2-G.

l (EOC)

I The limiting conditions for i

operation for the i

instrumentation that trips the recirculation pumps during turbine ~stop valve or control valve fast closure for transient margin improvement (especially for end of cycle) are given in Table 3.2-G.

H.

Accident Monitoring H.

Accident Monitoring Instrumentation

-Instrumentation The limiting conditions for Instrumentation shall be l

operation for'the accident.

calibrated and checked.as l

monitoring instrumentation are' indicated'in Table '4.2-H in all given in Table 3.2-H.

cperational modes other than COLD SHUTDOWN or refueling.

3.2-4 Amendment No. 109-I

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TIRE 4.2-0 a

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MINIMM TEST A0 CA.IIRATI(M FREQLDCY RR RM)lATI(M MMIT(RI!G SYSTE!6 5*.

Irstnnent Fisxtional Instnment Instrtment Ownnels Test (9)

Calibration (9)

Sute (heck Owxk

1) Refuel Area Exhast Pbnitors (hoe /3 nonths Ore / Refueling Oxe/nonth (hce/ day

2) Reactor Building Area Exhast Pbnitors Oxe/3 nonths (h / Refueling Oxm/nonth (hce/ day

3) Offgas Post-treatnent Radiation Ibnitors (hoe /3 nonths (10)

Oxe/ Refueling (hae/nonth Gre/ day wb

4) Offgas Pre-treatnent Radiation Pbnitors (h /3 nonths (10)

Ore / Refueling Oxm/nonth (hoe / day logic Systan Functional Test (4) (6)

Frequency (9)

1) Reactor Building Isolation (hce/ Refueling

2) Stanty Gas Treatnent Systen Actuation

-(hoe / Refueling

3) Stean Jet Air f.!ector Offgas 1.ine Isolation Ore /Refteling I

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DAEC-1 These instrument channels will be calibrated using simulated electrical signals.

4.

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

5.

Reactor low water level, high drywell pressure and high radiation main steam line tunnel are also included on Table 4.1-2.

6.

The logic system f'unctional tests shall include a calibration of time delay relays and timers necessary for proper functioning of the trip systems.

7.

These signals are not PCIS trip signals but isolate the Reactor Water Cleanup system only.

8.

This instrumentation is excepted from the functional test definition.

The functional test will consist of comparing the analog signal of the' active thermocouple element feeding the isolation logic to a redundant thermocouple element.

1 9.

Functional tests and calibrations are not required on the part of the system that is not required to be operable or is tripped.

Functional tests shall be performed prior to returning the system to an operable status with a frequency not less than once per month. Calibrations shall be performed prior to returning the system to an operable status with a frequency not less than those defined in the applicable table.

However, if maintenance has been performed on those components, functional tests and calibration shall be performed prior to returning to service.

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10. The Instrument Functional Test shall also demonstrate that control room '

alarm annunciation occurs if any of the following conditions exist:

1.

Instrument indicates measured levels above the alarm / trip setpoint.

2.

Instrument indicates a downscale failure.

3.

Instrument controls not set in operate mode.

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Amendment No. 109 i

DAEC-1 3.2.0.1 BASES 1.

Main Condenser Offgas Restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment.

This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50.

3.2-45b Amendment No.109 r

DAEC I

' LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT F.. Mechanical Vacuum Pump F.

Mechanical Vacuum Pump

-1.

The mechanical vacuum pump 1.

At least once during each shall be capable of being operating cycle verity isolated and secured on a automatic securing and signal of high radioactivity isolatior, of the mechanical in the steam lines whenever vacuum pump, the main steam isolation valves are open.

2.

During mechanical vacuum pump operation the release rate of gross activity except for halogens and particulates with half lives longer than eight days shall not exceed 1 curie /sec.

3.

If the limits of 3.7.F.2 are not met the vacuum pump shall be isolated.

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3.7-19b i

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Amendment No. 109 l

DAEC-1 i

3.7.F & 4.7.F Bases Mechanical Vacuum Pump The purpose of isolating the mechanical vacuum pump line is to limit the release of activity from the main condenser. During an accident, fission products could be transported from the reactor through the main steam lines to

. the condenser. The fission product radioactivity would be sensed by the main steam line radioactivity monitors which initiate isolation.

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3.7-48b Amendment No.109

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DAEC-1 i

LIMITING CONDITIONS FOR OPERATION-SURVEILLANCE REQUIREMENT 3.14 RADI0 ACTIVE LIQUID EFFLUENT 3.14.1 The radioactive liquid 4.14.1.1 Each radioactive liquid i

effluent monitoring effluent monitoring instrumentation i

instrumentation channels shown in channel shall be demonstrated Table 3.14-1 shall be OPERABLE OPERABLE by performance of the i

j with their alarm and trip CHANNEL CHECK, SOURCE CHECK, CHANNEL setpoints set to ensure that the CALIBRATION, and CHANNEL FUNCTIONAL limits of. Specification 3.14.2 are TEST operations during the modes and not exceeded.

at the frequencies shown in Table 4.14-1.

i As shown in Table I

APPLICABILITY:

I 3.14-1.

4.14.1.2 The setpoints shall be determined in accordance with the ACTION:

method described in the ODAM.

a. Een a radioactive liquid i

effluent monitoring instrumentation channel alann and trip setpoint is less conservative than a j

value which will ensure l

that the limits of 3.14.2 are met, adjust without i

delay to meet Specification 3.14.2, I

suspend the release of radioactive liquid i

i effluents monitored by the affected channel, or declare the channel l

inoperable.

b. When less than the minimum required liquid of fluent monitoring instrument channel is OPERABLE, take L

the ACTION stated in Table l

3.14-1 and make every l-reasonable effort to I

restore the instrument to operable-status.

In the I

event the minimum required instrumentation is not returned to OPERABLE status within 30 days, explain in the next Semiannual Radio' active Material Release Report, in lieu of any other report, why the instrument

.was not made OPERABLE in a timely manner.

4 Amendment No. 109.

3.14-1 l

DAEC-1 LIMITING CONDITIONS FCi?. OPERATION SURVEILLANCE REQUIREMENT 3.14.2 The concentration of 4.14.2.1 Each radioactive liquid radioactive material in liquid waste batch shall be sampled and effluent released from the site to analyzed in accordance with Table the unrestricted area (see FSAR 4.14-2 before release.

Figure 1.5-1) shall'not exceed the concentrations specified in 10CFR Alternatively, pre-release analysis Part 20, Appendix B, Table II, of batch (es) of radioactive liquid Colume 2 for radionuclides other waste may be by gross 6 or Y than dissolved or entrained noble counting provided the maximum gases. For dissolved or entrained permissible concentration, 1 x 10 7 noble gases, the concentration WCi/ml, is applied at the shall not exceed 2 x 10 " VCi/ml unrestricted area boundary.

total activity.

4.14.2.2 The results of pre-release APPLICABILITY: At all times, analyses shall be used with the calculational methods in the ODAM to ACTION:

establish trip setpoints for batch releases to assure that the

a. With the concentration of concentration at the restricted area radioactive material boundary does not exceed the limit released from the site to in Specification 3.14.2.

unrestricted areas exceeding the limit, 4.14.2.3 In any week during which without delay' restore the Service Water is released to the concentration within the unrestricted area, a grab sanple of limit.

water shall be collected from that Service Water System and analyzed as specified in Table 4.14-2, Item B.1 and B.4.

In the event the gross activity concentration in,the service water exceeds 1 x 10 - UCi/ml,the activity concentration shall be determined by sampling and post-release analyses specified in. Table 4.14-2, Items B.2 through B.5.

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3.14-2 1

Amendment No. 1.09 --

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RCQUIREMENT 3.14.3 The dose or dose commitment 4.14.3 Dose Calculations.

In any to a member of the Public from quarter in which radioacfive liquid radioactive materials in liquid effluent is discharged, an effluents released to the assessment shall be performed in unrestricted area (see FSAR Figure accordance with the ODAM at least 1.2-1) shall not exceed:

once per 30 days in order to verify that the cumulative dose commitment 1.5 mrem to the total body during does not exceed the limits in any calendar quarter, Specification 3.14.3.

5.0 mrem to any organ during any calendar quarter, 3.0 mrem to t'he total body during any calendar year, or 10.0 mrem to any organ during any calendar year.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding the above limit, prepare and submit to the Commission within 30 days from the end of the quarter during which the release occurred, pursuant to i

Specification 6.11.3, and in lieu of any other report, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken.

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/wendment No.109

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 3.14.4 Appropriate liquid radwaste 4.14.4 Each radioactive liquid waste equipment shall be used to treat batch shall be sampled and analyzed any untreated batch of liquid waste in accordance with Table 4.14-2 prior to discharge when a pre-before release, release analysis indicates a radioactivity concentration Alternatively, pre-release analysis (exclusive of tritium and dissolved of batch (es) of radioactive liquid noble gases) of 0.01 uCi/ml or waste may be by gross 8 or Y counting

higher, providedthemaximump9rmissible concentration, 1 x 10- uCi/ml,is APPLICABILITY:

At all times, applied at the unrestricted area boundary.

ACTION:

a. With radioactive liquid waste being discharged without treatment and in excess of'the above limit, prepare and submit to the Commission within 30 days, pursuant to Specification 6.11.3, a Special Report, in lieu of any other report, which includes the following information:

1 1.

Identification of equipment or subsystems i

not OPERABLE and the reason for inoperability.

2.

Action (s) taken to restore the inoperable equipment to OPERABLE status.

3.

Sunnary description of action (s) taken to prevent a recurrence, i

f 3.14-4 l

Amendment No 109

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I TABLE 3.14-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION

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MINIMUM CHANNELS.

i' INSTRUMENT OPERABLE APPLICABILITY #

ACTION l

1.

Gross Radioactivity Monitors

- Providing Automatic Termination of Release a.

Liquid Radwaste Effluent Line (1)

During releases 18 2.

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Gross Radioactivity Monitors Not Providing Automatic Termination of Release

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a.

RHR Service Water System Effluent Line (1)

During releases 20 j

b.

General Service Water System (1)

During releases 20 c.

RHR Rupture Disc Effluent (1)

During releases 20 Line 3.

Flow Rate Measurement Devices **

)

4 a.

Liquid Radwaste Effluent Line**

(1)

At all times 21 b.

Liquid Radwaste Dilution Line**

(1)

During releases 21 1

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1. Channel (s)'shall be OPERABLE and in service except that channels out of I

service are permitted for maintenance and required tests,' checks, or j

calibrations.

• • Pump curves may be utilized to estimate flow in lieu of flow measurement i

devices.

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f 3.14 5 Amendment No.109 l

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TABLE 3.14-1 (Continued)

TABLE NOTATION ACTION 18 With no channel OPERABLE, effluent may be released provided that prior to initiating a release:

1.

At least two samples are analyzed in accordance with Specification 4.14.2.1, and; i

2.

A technically qualified member of the Facility Staff verifies the release rate calculations and discharge valving determined by another technically qualified Facility Staff menber.

4 Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 20 With no channel OPERABLE, effluent releases via the affected l

pathway may continue provided the effluent is sampled and analyzed for gross radioactivity at least once per eight hours during actua1 release. The analysis shall be capable of detecti(g 10 7 pCi/ml.

ACTION 21 With no channel OPERABLE, effluent releases via this pathway may continue provided the flow rate is estimated with pump curves at

.least once per batch during actual releases.

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j 3.14-6 Amendment !!o.109

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TABLE 4.14-1 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 8

CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST 1.

Gross Beta or Gamma Radioactivity Monitors Providing Alarm and Automatic Isolation a.

Liquid Radwaste Effluent Line D*

D(6)

R(3)

Q(1) 24 2.

Gross Beta or Gamma Radioactivity Monitors Providing Alarm But Not Providing Automatic Isolation a.

RHR Service Water System Effluent Line D*

M R(3)

Q(2) b.

General Service Water D*

M R(3)

Q(2)

System Effluent Line c.

RHR Rupture Disc Effluent D*

M R(3)

Q(2)

Line 3.

Flow Rate Measurement Devices a.

Liquid Radwaste Effluent Line D(5)*

N.A.

R Q

b.

Liquid Radwaste Dilution Line D(5)*

N.A.

R Q

e e

9

TABLE 4.14-1 (Continued)

TABLE NOTATION

• During releases via this pathway.

• • During liquid additions to the tank.

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exist:

1.

Instrument indicates measured levels above the alarm / trip setpoint.

2.

Circuit failure.

3.

Instrument indicates a downscale failure.

4.

Instrument controls not set in operate mode.

(2) The CHANNEL FdNCTIONAL TEST shall also demonstrate that control room

~

alarm annunciation occurs if any of the following conditions exist:

i i

1.

Instrument indicates measured levels above the alarm / trip setpoint.

2.

Circuit failure.

l 3.

Instrument indicates a downscale failure.

j 4.

Instrument controls not set in operate mode.

l (3) The CHANNEL CALIBRATION shall include the use of a known radioactive source (traceable to the National Bureau of Standards radiation measurement system or acceptable non-NBS standards) positioned in a reproducible geometry with respect to the sensor and emitting beta or

~

gamma radiation in the range measured by the channel. CHANNEL CALIBRATION may normally be done during refueling outages.

(5) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once daily on i

i any dy on which continuous, periodic, or batch releases are made.

l (6) On any day on which a release is made, a SOURCE DIECK shall be made at least once, prior to the first release.

4 l

i i

3.14-8 l

A9eneent No.109

,_. ~

]{

]

TABLE 4.14-2

]

RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM h

Lower Limit

'~

=-

Minimum of Detection 5

Sampling Analysis Type of Activity (LLD)a g

Liquid Release Type Frequency Frequency Analysis (pC1/ml) 4 0

P P

8 A. Batch Waste Each Batch Each Batch Principal Gausna Emitters

• 5 x 10 7 i

Release Tanksc 1

i I-131' 1 x 10 6 i

1 P

f One Batch /M M

Dissolved and 1 x 10 5 Entrained Gases i

P i

Each Batch H-3 1 x 10 5 f

M b

I Composite Gross alpha 1 x 10 7 -

.w p

E Each Batch Sr-89, Sr-90 5 x 10.s Q'

b Composite Fe-55 1 x 10 6 f

B. Continuous 1.

W W

Gross beta /gansna 1 x 10 7 i

Service Water Grab Sample d

Release f

2.

.W W

Principal Gamma Emitters 5 x 10 7 4

Grab Sample I-131 1 x 10 6 f

i 3.

.M M

Dissolved and l

Grab Sample Entrained Gases 1 x 10 5 l

4.

W Mf H-3_

1 x 10 6 Grab Sample Composite i

l Gross alpha 1 x 10 7 f

l 5.

W Q

Sr-89, Sr-90 5 x 10 8 i

' Grab, Sample Composite n 55 1 - in_6 _

l l

i

4 s

TA8LE 4.14-2 (Continued)

TABLE NOTATION a.

The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a new i

count, above system background, that will be detected with 95% probability

. ith only 55 probability of falsely concluding that a blank observation i

w represents a "real" signal.

For a particular measurement, which may include radiochemical separation:

l 4.66 Sb LLD =

E

• V

• 2.22

• Y

• exp ( A A t) i I

where i

LLO is the lower limit of detection as defined above (picocuries per unit mass or volume) 1 Sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute)

Eisthecountingefficiency(countsperdisintegration) j V is the sample size (units of mass or volume)

[

2.22 is the number of disintegrations rer minute per picocurie, I

Y is the fractional radiochemical yield, when applicable, l

A is the radioactive decay constant for the particular radionuclide,.and' i

a t for effluents is the elapsed time between the midpoint of sample l

collection and the time of counting.

i Alternatively, exp ( A A t) may be replaced by_

A tt x exp-( At )

l 2

1 - exp(-At )

i Where:

ti is the total sampling time or sample compositing time i

t is the elapsed time between the end of sample collection and the time 2

j of counting.'

t Analyses shall be performed in such-a manner that the stated LLDs will be 1

achieved under routine conditions with typical valves of E, V,.Y and At for theradionuclidesMn-54,Fe-59,Co-58,Co-60,Zn-65,Mo-99,Cs-1$4,Cs-137, 1

Ce-141, and Ce-144. Occasionally background fluctuations, unavoidably small

}

sample sizes, interfering radionuclides, or other uncontrollable circumstances may render these LLDs unachievable.

i 3.14-10 1

A=n h nt No. 109

TABLE 4.14-2 (Continued)

~

TABLE NOTATION 4

When calculating the LLD for a radionuclide determined by ganna ray spectrometry, the background may include the typical contributions of other i

radionuclices 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.

The principal gama emitters for which the LLD specification will apply are F

exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134,-Cs-137, Ce-141, and Ce-144.

This list does not mean that only 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 LLD for the analyses should not be reported as being present at the LLD level. When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the~ Semiannual Radioactive Material Release Report.

b.

A composite sample is one in which the quantity of liquid sampled is j

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.

I c.

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

Before sanpling for analysis, each batch should be thoroughly mixed.

l d.

A continuous release is the discharge of liquid of a nondiscrete volume, e.g., from a volume of a system that has an input flow during j

the continuous release, e.

In the event a gross 8 or Y analysis is perfomed in lieu of an isotopic analysis before a batch is discharged, a sample shall be j

analyzed for principal gama emitters afterward.

f.

Analysis may be performed after release, t

i I

I 3.14-11 Amendment No. 109

DAEC-1 3.14.1 and 4.14.1 BASES

-1.

Radioactive Liquid Effluent Instrumentat-The radioactive liquid effluent instrumen..cion is provided to monitor 1

and control, as applicable, the release of radioactive material in liquid effluents. The OPERABILITY and use of these instruments l

implements the requirements of 10CFR Part 50, Appendix A, General Design Criteria 60, 63, and 64. The alann and/or trip setpoints for these instruments are calculated in the manner described in the 00AM to assure that the alarm and/or trip will occur before the limit specified in 10CFR Part 20.106 is exceeded.

Instrumentation is expected to be OPERABLE and in service when required by Specification. An instrument-may be removed from service voluntarily for the purpose of tests, checks, calibration, or preventative maintenance without declaring the channel inoperable, f

3.14.2 and 4.14.2. BASES i

1.

Liquid Effluent Concentration I

Specification 3/4.14.2 is provided to satisfy the regulation governing the maximum concentration of radioactive material in liquid effluent that may be released to an unrestricted area as stated in 10CFR Part 20.106 and the regulation requiring surveys needed to determine j

compliance stated in Part 20.201.

f Conformance to Specification 3.14.2, when applied to the activity concentration in the river at the site boundary due to liquid effluent, j

would assure 'that the average ~ activity. concentration in liquid effluent released to the unrestricted area is a small fraction of the limit specified in Part 20.106.

l 3.14.3 and 4.14.3 BASES 1.

Dose Due to Radioactive Effluents Specifications 3.14.3, 3.15.3, and 3.15.4 implement the requirements of 10CFR Part 50.36a and of 10 CFR Part 50, Appendix I, Section IV. These l

specifications state Limiting Conditions of Operation (LCO) to keep levels of radioactive materials in LWR effluents as low as is reasonably achievable. Compliance with these specifications will also keep average releases of radioactive material in effluents at small I

percentages of the limits specified in 10CFR Part 20.106. Surveillance I

requirements provide for the' measurement of releases and calculation of

. doses to verify compliance with the Specifications. Action statements in these Specifications implement the requirements of 10CFR Part 50.36(c)(2)'and 10CFR Part 50, Appendix I, Section IV.A in the event a LCO is not met.

l 3.14-12 l

i A Wnt No.- 109

_==_,

l

DAEC-1 2.

-Liquid Effluents With the implementation of Specification 3.14.3, there is reasonable assurance that Station operation will not cause a radionuclide concentration in public drinking water taken from the River that exceeds the standard for anthropogenic radioactivity in comunity drinking water. The equations in the ODAM for calculating doses due to measured releases of radioactive material in liquid effluent will be consistent with the methodology in Regulatory Guide 1.109 and 1.113.

The assessment of personal doses will examine potential exposure pathways including, as appropriate, consumption of fish and water taken from the River downstream of the discharge canal.

3.14.4 and 4.14.4 BASES 1.

Liquid Waste Treatment i

This specification implements the requirements of 10CFR Part 50.36a (a)

(1) that operating procedures be established and followed and that equipment be maintained and used to keep releases to the environment as low as is reasonably achievable. The specification intends that appropriate portions of the system which were used to establish compliance with the design objectives in 10CFR Part 50, Appendix I, Section II be used when specified to provide reasonable assurance that releases of radioactive material in liquid effluent will be kept as low as is reasonably achievable. The components in the liquid radwaste system which are appropriate to process liquid waste in order to i

satisfy Specification 4.14.4 are the floor drain demineralizer and the radwaste demineralizer.

The activity concentration, 0.01uci/ml, below which liquid radwaste treatment would not be' cost-beneficial, and therefore not required, is demonstrated below.

The quantity of radioactive material in liquid e{ fluent released annually from the DAEC has been calculated tn be total iodines 0.11 curie 3

total others (less H )

0.25 Total ETI curie i

The population dose comitment resulting from the radioactive material l

in liquid effluent released annually has been calculated to be thyroid 0.164 man rem total body 0.114 Total E77K man rem I" Evaluation of the Duane Arnold Energy Center to demonstrate Conformance to the Design Objectives of 10 CFR 50, Appendix I, " Iowa Electric Light & Power Company, May 1976.

I s

3.14-13 l

Amendment No.109

DAEC-1 Therefore, population' doses are about 1.5 man rem per curie of iodine 3

released and about 0.5 man rem per curie of other radionuclides (less H )

released-in liquids. On the basis of gross activity, the population dose is about one man rem per curie released in liquids.

The volme of liquid waste processed and intended for discharge is estimated to be:

6 Low Purity Waste 5700 gal / day = 1.8 x 10 gal /yr 5

Chemical Waste 600 gal / day = 1.9 x 10 gal /yr Since the same DAEC equipment is used to process both streams, the total 6

volume to be processed is about 2 x 10 gal /yr.

The annual cost to operate the radwaste processing equipemnt, based on Dirty 2

Waste Ion Exchange operation, has been estimated (neglectingcreditfor capital recovery) to be $88000 per year.

Thus the unit volume operating cost-is about:

$88000/yr

= $0.05/ gal 2 x 10 gal /yr Thus the operating cost to treat a 4000 gallon batch of chemical waste by ion exchange would be about $200. The operating cost to treat a 10000 gallon batch of floor drain waste by ion exchange would be about $500.

Assuming the cost-benefit balance is $1000 expenditure per man rem reduced and assuming treatment removes all radioactivity from the liquid, then (1) the activity concentration in a Chemical Waste batch below which treatm6nt is not cost-beneficial is 6

C=

$200 x 1 curie x 10 u Ci x 1 man rem 4000 gal x 3785 m1 man rem curie 51000 gal C = 0.013 9Ci/ml (2) the activity concentration in a batch of Floor Drain Waste below which treatment is not cost-beneficial is 6

C=

$500 x 1 curie x 10 u Ci x 1 man rem 10000 gal x 3785 m1 man rem curie 51000 gal C = 0.013 uCi/ml 2 Ibid., based on Regulatory Guide 1.110 3.14-14 Amendment No.109

DAEC-1 Liquid waste treatment with the evaporator at DAEC has been shown to be neither cost-beneficial nor necessary to comply with 10CFR50 Appendix I, Section II design objectives.

Consequently, liquid radwaste treatment to achieve an activity concentration below 0.01 uCi/mi in liquid effluent is not justified.

)

i i

i e

3.14-15 l

4 Amendment No. 109

DAEC-1

_ LIMITING CONDITIONS FOR OPERATION SURVEILLANCEREQUkREMENT 3.15.1 The radioactive gaseous 4.15.1.1 Each radioactive gaseous effluent monitoring instrumentation effluent monitoring instrumentation channels shown in Table 3.15-1 channel shall be demonstrated shall be OPERABLE with noble gas OPERABLE by performance of the alarm setpoints set to cause CHANNEL CHECK, SOURCE CHECK, CHANNEL automatic. alarm when the limits of CALIBRATION, AND CHANNEL FUNCTIONAL Specification 3.15.2.1 are TEST operations during the MODES and exceeded.

at the frequencies shown in Table 4.15-1.

APPLICABILITY: As shown in Table 3.15-1.

4.15.1.2 The setpoints shall be determined according to the method ACTION:

described in the ODAM.

a. With radioactive gaseous effluent monitoring instrumentation channel alarm setpoint less conservative than a value which will ensure that the limits of 3.15.2 are met, i

adjust without delay to meet Specification 3.15.1, declare the channel inoperable, or immediately suspend any release via the instrumented pathway.

i

b. When less than the minimum required gaseous effluent monitoring instrument channels OPERABLE, take the sction stated in Table 3.15-1 and make every reasonable effort to restore the instrument to operable status.

In the event the minimum required instrumentation is not l

returned to OPERABLE status within 30 days, explain in the next Semiannaul Radioactive Material Release Report, in lieu of any other report, why the instrument was not made OPERABLE in a timely manner.

.s 3.15-1 L

Amendemnt No.109

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 4.15.2 Compliance with 3.15.2 shall be assessed on the basis of results of measurements specified in Table 4.15-2 and according to methodology stated in the 00AM.

3.15.2.1 The dose rate in the unrestricted area (see FSAR Figure 1.5-1) due to radioactive noble gas-released in effluents shall not exceed 500 mrem / year to the total j

body or 3000 mrem / year to skin.

3.15.2.2 The dose rate in the unrestricted area due to I-131, I-133, H-3, and to radioactive particulates hav.ing half-lives of 8 days or more that are released in effluents shall not exceed 1500 mrem / year to any organ.

APPLICABILITY: Whenever monitoring or sampling is required.

ACTION: When the dose rate exceeds a limit in 3.15.2, decrease the i

release rate without delay to comply with the limit.

3.15-2 l

Amendment No. 109

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 3.15.3 The air dose in the 4.15.3.1 Dose Assessment An 4

unrestricted area (see UFSAR Figure assessment shall be performed in 1.2-1) due to noble gases released accord with the 00AM at least once in gaseous effluents shall not every 30 days to verify that the exceed:

cumulative air dose during the quarter and year due to noble gases 5.0 mrad from gamma radiation does not exceed the limits in i

during any calendar quarter, Specification 3.15.3.

10.0 mrad from beta radiation i

during any calendar quarter,

}

10.0 mrad from gamma radiation during any calendar year, or, 20.0 mrad from beta radiation l

during any calendar year.

APPLICABILITY: "At all times when

~

l monitors are required.

l ACTION:

a. If the calculated air dose i

l from radioactive noble gases in gaseous effluents exceeds either of the above limits prepare and submit a Special Report to the i

Comission within 30 days following the end of the i

calendar quarter during which the release occurred.

The Special Report shall be pursuant to Specification 6.11.3, shall be ir, lieu of any other report, and shall l

identify the cause(s) for exceeding the limit and define the corrective l

actions taken.

I l

..o 3.15-3 l

Amendment No.109 l

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 3.15.4 -The dose to a member of the 4.15.4.1 Dose Assessment An' 1

public from iodine-131, I-133, H-3, assessment shall be perfonned in l

and from radionuclides in accordance with the 00AM at least l

particulate form having half-lives once every 31 days to verify that greater than eight days in gaseous the cumulative dose commitment due effluents released from the site to to I-131,1-133 H-3, and the unrestricted area (see FSAR radioactive particulates having Figure 1.5-1) shall not exceed:

. half-lives greater than eight days in gaseous effluents does not exceed 7.5 mrem to any organ during any the limits in Specification 3.15.4.

calendar quarter, or, 15.0 mrem to any organ during any calendar year.

I APPLICABILITY: At all times when monitors are required.

ACTION:

~

a. With the calculated dose from the release of I-131, I-133 H-3, and radionuclides in i

particulate form having half-lives greater than i

eight days in gaseous l

effluents exceeding the above limit, prepare and i

submit a Special Report to the Commission within 30 days following the end of the calendar quarter during which the release occurred.

The Special Report shall be made pursuant to l

Specification 6.11.3, shall be in lieu of any other report, and shall identify the cause(s) for exceeding i

the limit and define the j

corrective actions taken.

i 3.15-4

-l Amendment No. 109 l

l DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 3.15.5 Every reasonable effort 4.15.5 The gaseous effluent shall be made to maintain at least monitoring program of Specification one train of the Offgas System 3.15.1 shall be used to verify the OPERABLE.

operation of.the offgas system.

Within four hours after conmencing operation of the main condenser air ejector, at least one train of charcoal beds in the Offgas System shall be placed in operation to treat radioactive gases from the main condenser air ejector. During continuing reactor operation, at least one train of charcoal beds in the Offgas System shall be used to treat the gases before discharge.

APPLICABILITY:~ When the main

(

condenser air ejector is operating.

ACTION:~

j-l

a. If gaseous wastes are discharged for more than 7 days without treatment, prepare and submit a Special Report to the

+

Commission within 30 days pursuant to Specification 6.11.3, in lieu of any l

other report, including the following information:

i'

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

l

2. Action (s) taken to restore the inoperable l

equipment to OPERABLE status.

l

3. Summary description of j

action (s) taken to l

prevent a recurrence.

r 11F 3.15-5 Amendment No.'109

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT l

3.15.6 The concentration of 4.15.6 The concentration of hydrogen in the offgas system hydrogen in the Offgas System shall downstream of the recombiners shall be determined by monitoring the be limited to $_4% by volume.

offgases in the Offgas System downstream of the recombiners with APPLICABILITY: During Offgas the hydrogen monitors.

System operation.

ACTION:

a. With the concentration of hydrcgen in the main condensor offgas treatment system downstream of the recombiners exceeding the limit, restore the concentration to within the limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

b. In the event the hydrogen concentration is not reduced to < 4% within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, be in at least HOT SHUTDOWN or within the limit within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l l

l l

l l

3.15-6 Amendment No. 109

TABLE 3.15-1 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION j

Minimum Channels Instrument 9 Operable Applicability #

Function Action n

z 1.

Offgas Post-Treatment Noble Gas Activity 1

Monitor activit 25 Monitor (RI) concentration, yalann 2.

Offgas Hydrogen Monitor (R2) 2 Monitor _ hydrogen 29 concentration 3.

Offgas Stack Monitoring System (R3)

Monitor activity 27 a.

Noble Gas Activity Monitor 1

concentration alarm Collect iodine, sample 31 b.

Iodine Sampler Cartridge 1

Collect particulate 31 I

c.

Particulate Sampler Filter 1

sample d.

Effluent Flow Measuring Device 1

Measure air flow 26

' u, e.

Sample Flow Measuring hvice 1

Measure air flow 26 u,

la 4.

Reactor Building Exhaust Vent Monitoring System (R4?

a.

Noble Gas Activity Monitor 1

Monitor activity 27 concentration, alarm i

b.

Iodine Sampler Cartridge 1

Collect iodine sanple 31 l

Collect particulate 31 l

c.

Particulate Sampler Filter 1

sample d.

Effluent Flow Measuring Device 1

Measure air flow 26 e.

Sample Flow Measuring h vice 1

Measure air flow 26 (5. Turbine Building Exh3ust Vent Monitoring System (RS) a.

Noble Gas Activity Monitor 1

Monitor radioactivity 27 concentration, alarm b.

Iodine Sampler Cartridge 1

Collect iodine sample 32 Collect particulate sample 32 c.

Particulate Sampler Filter 1

d.

Effluent Flow Measuring Device 1

Measure air flow 26

{

e.

Sample Flow Measuring Device 1

Measure air flow 26 0

TABLE 3.15-1 (Continued)

TABLE NOTATION 9 Refer to ODAM Figure 3-1 for location of effluent monitoring points R1 thru R6.

1. Channels shall be OPERABLE and in service except that channels out of service are permitted for the purpose of required tests, checics, calibration, and preventative maintenance without declaring the channel to be inoperable.

• During releases via this pathway.

• • During main condenser offgas treatment system operation.

• • • During operation of the main condenser air ejector.

ACTION 25 With no channel OPERABLE, gases from the main condenser offgas treatment system may be released to the environment for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided:

1.

The offgas delay system is not bypassed; and 2.

The offgas stack noble gas activity 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 />.

ACTION 26 With no channel OPERABLE, effluent releases via this pathway may l

continue provided the flow rate is estimated whenever operation of a main exhaust-fan combination is changed in the system.

ACTION 27 With no channel OPERABLE, effluent releases via this pathway may continue if grab samples are taken at least once per eight hours and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or if an alternate monitoring system is utilized.

Drywell purge is permitted whenever the offgas stack monitor or its alternate monitor is operating.

ACTION 28 Deleted ACTION 29 With one channel OPERABLE operation of the main condenser offgas treatmentsystemmaycontlnueprovidedtherecombinertemperature sensor is operable. When only one of the preceeding methods is operable, the offgas system may be operated provided gas samples are collected at least once per day and analyzed for hydrogen within the ensuing four hours.

ACTION 31 With the no channel OPERABLE effluent releases via this pathway -

may continue, provided sample,s required in Table 4.15-2 are continuously collected with auxiliary sampling equipment, i

ACTION 32 With no channel OPERABLE, effluent releases via this pathway may l

continue. if grab sanples are taken at least once per eight hours and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> I

j or if an alternate monitoring system is utilized.

3.15-8 Amen ent No.109 dm

~

~ _

i TABLE 4.15-1 RADI0 ACTIVE GASEQUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

(

CHANNEL g

CHANNEL SOURCE CHANNEL FUNCTIONAL REQUIRED

=

INSTRUMENT CHECK CHECK CALIBRATION TEST MODE #

E I

1.

Offgas Hydrogen Monitor D**

N.A.

Q(4)

M m

2.

Offgas Stack Monitoring System a.

Noble Gas Activity Monitor D*

M R(3)

Q(2) b.

Iodine Sampler Cartridge W*

N.A.

N.A.

N.A.

c.

Particulate Sampler Filter W*

N.A.

N.A.

N.A.

d.

Effluent Flow Measuring Device D*

N.A.

R Q

e.

Sample Flow Measuring Device D*

N.A.

R Q

g 3.

Reactor Building Vent Monitoring System a.

Noble Gas Activity Monitor D*

M R(3)

Q(2) w b.

Iodine Sampler Cartridge W*

N.A.

N.A.

N.A.

T c.

Particulate Sampler Filter W*

N.A.

N.A.

N.A.

d.

Effluent Flow Measuring Device D*

N.A.

R Q

e.

Sample Flow Measuring Device D*

N.A.

R Q

4.

Turbine Building Exhaust Ventilation Monitoring System 4

a.

Noble Gas Activity Monitor D*

M R(3)

Q(2) b.

Iodine Sampler Cartridge W*

N.A.

N.A.

N.A.

c.

Particulate Sampler Cartridge W*

" A.

N.A.

N.A.

d.

Effluent Flow Rate Monitor D*

N.A.

R Q

e.

Sample Flow Measuring Device D*

N.A.

R Q

m 9

l TABLE 4.15-1 (Continued)

TABLE NOTATION i

1. . Instrumentation shall be OPERABLE and in service except that channels out of service are permitted for the purpose of required tests, checks, calibrations, and preventative maintenance without declaring the channel to be inoperable.

During releases via this pathway.

• • During main condenser offgas treatment system operation.

• • • During operation of the steam jet air ejector.

(1) Not.used.

f (2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:

1.

Instrument indicates measured levels above the alarm setpoint.

2.

Circuit failure.

3.

Instrument indicates a do.voscale failure.

4.

Instrument controls not set in operate mode.

9 (3) The CHANNEL CALIBRATION shall include the use of a known radioactive source (traceable to the National Bureau of Standards radiation measurement system or other acceptable non-NBS standards) positioned in a reproducible geometry with respect to the sensor and emitting beta and/or gamma radiation in the range measured by the channel in accord with established station calibration procedures. Alternately, after the initial calibration, noble gas activity monitors may be calibrated by laboratory analyzed gas samples collected and analyzed per Table 4.15-2, item A.

(4) The CHANNEL 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, balance nitrogen.

t 3.15-10 i-Amendment No. 109

TABLE 4.15-2 g

ilADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM h

Minimum Lower Limit of y

Sampling Analysis Type of Detection (LLD)a Gaseous Release Type Frequency Frequency Activity Analysis (pCi/ml) r+

D b

A. Offgas Stack, and M Grab Sample M

Principal Gama Emitters 1 x 10 " e g

Reactor Building Vent Q9 Grab Sample Q

H-3 1 x 10 6 d

B. Offgas Stack, Continuous yc I-131 1 x 10 12 Reactor Building Charcoal Vent, and Turbine Sample Building Vent d

Wc Principal Gama Emitters 1 x 10 11 e Continuous Particulate (I-131,Others)

Sample w

h Continuous Q

Sr-89, Sr-90 1 x 10 11 d

0 Composite Particulpte Gross Alpha 1 x 10 11 Sample C. Offgas Stack,

' Continuous Continuous Radioactive Noble Gas 1 x 10 6 Reactor Building gama activity Vent, and Turbine Building Vent S

O

I TABLE 4.15-2 (Continued)

TABLE NOTATION Table 4.14-2, Note a is a definition of the lower limit of detection a.

(LLD).

b.

Analyses shall be performed following an increase of more than 50% in the steady state releases as indicated by a noble gas activity monitor, i

after factoring out the effect due to a change in reactor power.

Sample media shall be changed at least once per seven days and the c.

analysis completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing (or after removal from the sampler).

Analyses shall.also be performed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following an increase of more than 50% in the steady state release as indicated by a noble gas activity monitor, after factoring out the effect due to a change in reactor power. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less are analyzed, the corresponding LLD may be increased by a factor of 10; d.

The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.15.2, 3.15.3 and 3.15.4.

The principal gama emitters for which the LLD specification will apply e.

are exclusively the following radionuclides: Kr-87, Kr-88, Xe-133, l

Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Hn-54, Fe-59, i

Co-58, Co-60, Zn-65,.Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for i

particulate emissions. This list does not mean that only 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 LLD may i

~

be reported as "less than" their respective LLD and should not be reported as being present at the LLD of the nuclide. Each measured radionuclide concentration is used in a required concentration or dose calculation only if it is detected at or above the LLD. When unusual 4

circumstances persist more than 30 days and cause LLD higher than required, the reasons shall be documented in the Semiannual Radioactive i

Material Release Report.

i f.

A quarterly composite sample shall include an equal fraction of each i

weekly particulate sample collected during the quarter.

An H-3 grab sample will also be taken'from the'0ffgas Stack or Reactor.

g.

Building Vent when the reactor head is removed.

L i

j.

r I

3.15-12.

Amendment No. 109

l DAEC-1 3.15.1 and 4.15.1 BASES 1.

Radioactive Gaseous Effluent Instrumentation The radioactive gaseous effluent instrumentation is provided to monitor the release of radioactive materials in gaseous effluents and, as appropriate, to control potential releases.

Instrurnentation for monitoring the concentration of potentially explosive gas mixtures in the main condenser offgas treatment system is also provided.

The presence of instruments for monitoring both radioactive and explosive gaseous effluents is depicted in ODAM Figure 3-1.

The OPERABILITY and use of these instruments implements the requirements of 10CFR Part 50, Appendix A, General Design Criteria 60, 63, and 64.

Offgas post-treatment monitors are operable during reactor power.

operation with their trip setting at a value not exceeding a limit computed by a method described in the Offsite Dose Assessment Manual.

If both instruments reach their high trip point or if one reaches the 4

high trip point and the other reaches a downscale trip point, the charcoal delay bed bypass valvas close immediately.

Reactor building exhaust ventflation shaft radiation monitors initiate isolation of the reactor building normal ventilation and start standby gas treatment when a high trip point is reached.

DAEC is equipped with a radi.oactive gaseous effluent monitoring system which includes detectors at the offgas stack (R3), the reactor building vent (R4), and the turbine building vent (RS). A remote indication and control unit (RIC) located. near each oetector displays the detector reading and, whenever 'the setpoint is exceeded, an indicator light.

The data are also routed to a control computer and a control room display but do not 'cause a trip to isolate the ventilated area.

In the event the control computer and/or control room display f ail to function or are voluntarily take1 out of service, it is intended that each affected RIC display br: observed at least once per hour (in which case the affected channel remains OPERABLE).

In the event the detector r

i reading and the indication of exceeding the monitor setpoint are not provided at either the control room or the RIC, then the affected channel is not OPERA 3LE and DAEC will either perform the appropriate ACTION or will provide an alternate monitoring system. This permits DAEC to retain the GE gaseous monitoring system as an alternate system for normal effluent monitoring when the Kaman system is temporarily inoperable. When used as an alternate monitoring system, the GE system is subject to the. requirements stated in Specifications 3.15.1 and 4.15.1 and to LLD requirements' stated in Table 4.15-2, Item.C.

3.

Gaseous Effluents Assessments of dose required by Specifications 4.15.3 and 4.15.4 to verify compliance with Appendix I,Section IV are based on measured radioactivity in gaseous effluent and on calculational methods stated i

in the ODAM. Pathways of exposure and location of individuals are selected such that the dose to a nearby resident is unlikely to be 3.15-13

' AmeWnt No.109 i

DAEC-1 underestimated.

Dose assessment methodology described in the ODAM for gaseous effluent will be consistent with the methodology in Regulatory Guides 1.109 and 1.111. Cumulative and projected assessments of dose made during a quarter are based on historical average, i.e.,

quarterly averaged conditions measured at DAEC. Assessment made for the annual radiological environmental report will be based on annual averages of atmospheric conditions during the period of release.

3.15.2 and 4.15.2 8ASES Gaseous Effluent Concentration This specification is intended to ensure that the concentration of radioactive material in the unrestricted area beyond the site boundary due to gaseous effluents from DAEC will maintain doses within the annual dose limits to unrestricted area provided in 10 CFR Part 20.

Compliance with these limits also reasonably assures that radioactive material in gaseous effluents will not result in exposure of a member of the public in an unrestricted area to annual avera'ged concentrations exceeding the limit in 10 CFR Part 20.106. The occupancy time of members of the public who may occasionally be on the site is expected to be low enough to compensate for any less atmospheric dispersion on site than to the environs offsite.

Assessment of compliance is based upon an effluents measurement program defined in Table 4.15-2 and methodology stated in the ODAM. The resolving time of the measurements, ie., the sample integration time, bounds the minimum averaging time of the effluent measurements. waste streams. The Standby Gas Treatment System is considered an Engineered Safety Feature and not an exhaust ventilation treatment system. Thus the exhaust ventilation system discharges via the reactor building vent.

i 3.15.3 and 4.15.3 BASES Doses due to Noble Gases These specifications implement the requirements of 10 CFR Part 50, Appendix I.

t 3.15.4 and 4.15.4 BASES Doses due to Iodine' and Particulates in Air These specifications implement 10 CFR Part 50, Appendix I.

The dose calculation methods in the ODAM depend on existing pathways of exposure to a member of the public or more conservative conditions assumed (yielding a higher calculated dose).

Calculations and methods are such that an estimate of the dose to a member of the public is not likely to be underestimated substantially.

3.15-14 Amendment No.109

DAEC-1 3.15.5 and 4.15.5 BASES 1.

Gaseous Radwaste Treatment This specification implements the requirement of 10 CFR Part 50.36a (a)(1) that operating procedures be established and followed and that equipment be maintained and used to keep releases to the environment as low as is reasonably achievable.

In order to satisfy Technical Specification 3.15.5, every reasonable effort shall be made to maintain and operate at least one train of the Offgas System charcoal adsorbers with pre-and aft-particulate filters to _ process radioactive gaseous effluent prior to release.

The specification that the Offgas System which was used to establish compliance with the design objectives in 10CFR Part 50, Appendix I, Section II be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept as low as is reasonably achievable.

ODAM Figure J-1 is a flow diagram depicting gaseous radioactive waste.

streams. The Standby Gas Treatment System is considered an Engineered Safety Feature and not an exhaust ventilation treatment system.

3.15.6 and 4.15.6 BASES 1.

Explosive Gas Mixture Specification 3/4.15.6 is provided to ensure that the concentration of potentially explosive gas in the offgas treatment sytem downsteam of the recombiners is maintained below the flammability limit of a hydrogen and oxygen mixture in the system. Keeping the mixture below its flammability limit will provide assurance that offgas treatment system integrity and operability is maintained and that the radioactive material concentration is the offgas will be controlled in conformance with 10CFRPart 50, Appendix A, Criterion 60. Calibration gas concentrations will be within the range of interest for hydrogen concentration and will not include 0% or 100% hydrogen concentrations.

Amendment No.'T09

~

i

DAEC-1 1

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 3.16.1 The annual dose or dose 4.16.1 Dose Calculations.

comitment to any member of the Cumulative dose contributions from public due to radiation and liquid and gaseous effluents to a radioactive material in effluents member of the public offsite shall from DAEC shall not exceed 75 mrem be evaluated at least once every to his thyroid or 25 mrem to his year as described in the ODAM.

total body or any other organ.

APPLICABILITY: At all times.

ACTION:

a.

If the calculated dose from radioactive material released in liquid or gaseous effluents exceeds twice the limits of Specifications 3.14.3, 3.15.3, or 3.15.4, perform an assessment of compliance with 10 CFR 190 and limit subsequent releases such that the dose or dose comitment to a member of the public is < 75 mrem to his thyroid and < 75 mrem to his total' body or any other organ over 12 consecutive months including the period of ele'vated release, b.

If the' estimated' dose exceeds either limit in Specification 3.16.1, prepare and submit a Special Report.to the NRC within 30 days in lieu of any other report; it shall include the cause of the release of exposure, an estimate of the dose to the-likely most exposed member (s) of the public, corrective actions taken or planned to prevent a recurrence, and a schedule for achieving compliance.

If th condition causing the limit (e) s to be exceeded has not been i

corrected the Special Report may also s, tate 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 gra.*ted until NRC action on the reque:t is complete.

3.16-1 l

- AmeWant No.109 -

=. _ _ -

=

DAEC-1 f

i LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 3.16.2 A radiological 4.16.2.1 Sampling and analyses environmental monitoring program required in Table 3.16-1 shall be shall be conducted as specified in performed such that the detection Table 3.16-1.

capabilities specified in Table 3.16-2 are achieved under routine APPLICABILITY:

At all times.

conditions.

If a sample analysis does not meet the LLD specified, ACTION:

report the reason attributed in the next Annual Radiological

a. In the event the Environmental Report, radiological environmental monitoring program is not 4.16.2.2 Land Use Census DAEC conducted as specified in shall conduct annually a land use the Table.3.16-1, prepare census within three miles of the and submit to the

. Station to identify radiologically Connission in the Annual important changes in land use.

Radiological Environmental l

Report the reasons for not conductin.g the program in i

accord with the Table 3.16-l 1 and the plans for preventing a recurrence.

b. In the event radioactivity in a sampled environmental medium, averaged over a calendar quarter, is 4

attributable to DAEC and exceeds an appropriate value listed in Table 3.16-3 or, if not listed, causes a potential annual dose exceeding two times the quarterly dose limit in Specification 3.14.3 or 3.15.4, prepare and submit to the Comission within 30 i

after discovery a Special Report which includes an evaluation of any release conditions, environmental factors or other conditions which caused the value(s) i of Table 3.16-3 or. two times the quarterly dose limit to be exceeded and which defines the corrective actions to be taken.

If the radioactivity in environmental sample (s) is not attributable to releases from the Station, l

the Special Report is not e

I

%nendment No.109

DAEC-1

-LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT l

required.

Instead, the-sample'(s) result (s) shall be reported and explained in the Annual Radiological Environmental Report.

c. When environmental sampling medium is not available from a sampling location or the location is no longer appropriate, the cause and the location where replacement samples were obtained and/or will be obtained shall be reported in the Annual Radiological Environmental Report.

d. In the event a location is identified at which the calculated personal dose associated with one or more exposure pathways exceeds by 20% the maximum calculated dose associated with like pathway (s) at a location where sampling is conducted as specified by Table 3.16-1, then the pathway (s) having maximum exposure potential at the newly identified location will be added to the' radiological monitoring program at a subsequent Operations Committee meeting, if samples are reasonably attainable at the new location.

Like oathway(s) monitored l

'(sampled) at a location,

~

excluding the control station location (s), having a lesser associated calculated personal dose may be deleted from the program at the time the new pathway (s) and location are added.

3.16-3 l

Amenhent No.109

DAEC-1 i

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 3.16.3 Analyses shall be performed on radioactive materials supplied in an Interlaboratory Comparison 1 Program which has been approved by the NRC.

APPLICABILITY:. Applicable to the Radiological Environmental Monitoring Program at all times.

. ACTION:

In the event analyses were not performed as required in Specification 3.16.3, report the corrective actions taken to prevent a recurrence in the Annual Radiological Environmental Monitoring Report.

l i

I l

l l.

3.16-4 imendment No.109 I

DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT

,i 3.16.4 Appropriate equipment shall 4.16.4.1 The Process Control be operated in accordance with a Program shall state the essential Process Control Program to process operating parameters of the wet radioactive waste solids process (es), the essential destined for disposal Jto a form characteristics of the waste form to that meets. appropriate requirements be shipped, and the essential of 10 CFR Part 61.56 before the product verification requirements.

waste is shipped from the DAEC site.

4.16.4.2 Before a Contractor processes radioactive waste, DAEC APPLICABILITY: During Processing shall verify that he has an NRC of radioactive waste solids for approved Process Control Program, disposal.

ACTION:

1.

Suspend delivery to a transport carrier of any container of radioactive wasti not complying with 10 CFR Part 61.56.

i i

3.16-5 A=andment No. 109 gy I

i

TABLE 3.16-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 1

A Exposure Pathway Minimum Number of Sampling and and/or Sampling Stations Collection Frequency Type and Frequency of Analysis 2y Sample Type 5

Airborne five Continuous operation of

• Analyze for gross beta activity > 24 Particulates sampler with sample hours after filter change.

collection at least

• Perform gamma isotopic analysis on once per week or as each sample having gross beta required by dust loading activity > 10 times the yearly mean of control samples.

• Perform gama isotopic analysis on composite (by sampling location) of samples collected during each quarter Airborne five Continuous operation of Analyze each cartridge for I-131.

" Radioiodine sampler with sample

.g collection at least once 4

per week.

Ambient thirty-eight Two dosimeters at each Read gamma radiation dose quarterly.

Radiation point continuously.

Change at least once per quarter.

Surface Water two At least once per month.

• Gamma isotopic analysis of each sample or monthly composite (by location).

• Tritium analysis of a composite (by location) at least once per quarter.

Ground Water four At least once per Analyze quarterly for tritium and (potable) quarter.

(May be gross beta activity if gross beta composited if collected

>10timestheyearlymeanofcontrol more frequently.)

samples, analyze for SR-89, SR-90, l

and gama isotopic.

River Sediment one At least once every six Gama isotopic analysis of each months sample.

l

TABLE 3.16-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Minimum Number of Sampling and and/or Sampling Stations Collection Frequency Type and Frequency of Analysis Sample Type Milk four At least once per two Gamma isotpic and I-131 analysis of weeks (biweekly) during each sample, the grazing season. At least once per month during non-grazing season.-

Fish two Two times per year.

Ganna isotopic analysis on edible w

(Once during January portion.

?

through July and once during Au ust thru December.

Vegetation three-Annually at harvest time.

Gamma isotopic analysis of edible One smple of each:

portion.

grain

(

green leafy vegetation forage.

one One sample of broadleaf I-131 analysis vegetation at time of harvest Required smple station locations are described in the Offsite Dose Assessment Manual.

TABLE 3.16-2 i

MAXIMUM VALUES OF THE LOWER LIMIT OF DETECTION FOR ENVIRONMENTAL SAMPLE ANALYSISa Medium 3p Airborne Particulate Water orGag Fish Milk Food Products Sediment s

Analysis (pCi/1)

(pCi/m )

(pCi/kg, wet)

(pCi/1)

(pC1/kg, wet)

(pCi/kg, dry) 2 2"

a gross beta 4

1 x 10 2 e

3 b

H 2000 3000c 5"Mn 15 130 58Fe 30 260 58 68 00, 00 15 130 w

65Zn 30 260 m

85

. ln Zr 30 4-95Nb 15 1

500c 7 x 10 2 1

60-131 13"Cs 15 5 x 10 2 130 15 60 150 137Cs 18 6 x 10 2 150 18 80 180 l

I"08a 60 60 I"0La 15' 15 4

e i

DAEC-1 TABLE 3.16-2 (Continued)

TABLE NOTATION l'

a.

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

For a particular measurement, which may include radiochemical separation:

1. ' " *b uD-E ? Y

• 2.22

• Y

• exp ( A a t) where LLD is the lower limit of detection as defined above (picocuries per unit mass or volume) l Sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute)

I E is the counting efficiency (counts per disintegration)

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

2.22 is the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yield, when applicable, A is the radioactive decay constant for the particular radionuclide, and A t 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 the stated LLDs will be achieved under routine conditions. With typical values of E, V, Y, and At for the radionuclides named in the Table. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable.

In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report.

When a radionuclide attributable to DAEC but not listed in this table is measured (more than the LLD) ld not be reported as being present at the LLD it shall be reported.

Any nuclide that is below the LLD for the analysis shou l

level, b.

For Drinking Water.

c.

For samples of water not used as a source of drinking water, w

3.16-9

• .. endment No.109 l

.. ~ _ _.

TABLE 3.16-3 g

REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES _

k Reporting Levels (a) o Water AirborneParticujate Fish Milk Food Products g

Analysis

-(pC1/1) or Gases (pCi/m )

(pC1/Kg, wet)

(pCi/1)

(pC1/Kg, wet) h H-3 2 x 10 (b?

4 3 x 10%cJ 3

Mn-54 1 x 10 3 x 10"

?

Fe-59 4 x 10 1 x 10" 3

Co-58 1 x.10 3 x 10" w

2 h

Co-60 3 x 10 1 x 10" h

Zn-65 3 x 10 2 x 10" 2

Zr-Nb-95 4 x 10 (c) 2 I-131 2(c) 0.9 3

1 x 10 2 3

3 Cs-134 30 10 1 x 10 60 1 x 10 3

3 Cs-137 50 20 2 x 10 70 2 x 10 Ba-L a-140 2 x 10 (d) 3 x 10 (d) 2 2

(a) - The reporting level is exceeded when one or more radionuclides is detected in a sample and concentration (1)_

concentration (2). # *

• y i*

+

reporting sevel (1) reporting leves (z)

(b) - For drinking water samples.

This is 40CFR Part 141 value.

(c) - For samples of water not used as a source of drinking water.

(d) --Concentration of parent or, daughter.

DAEC-1

-3.16.1 and 4.16.1. BASES 1.

Dose i

Specification 3.16.1 is provided to comply with the dose limitation requirement of 40CFR190. This specification requires the assessment of dose to demonstrate that a person (a nearby resident) has not received a radiation dose exceeding that specified in 40CFR190. including doses from direct radiation.

There is no other licensed nuclear fuel cycle facility within 50 miles of DAEC, thus it is assumed that the dose from

)

other uranium fuel cycle facilities is negligible.

In the event a report is required to satisfy Specification 3.16.1, Action b, it shall be deemed adequate to satisfy the reporting requirement in Specification 6.11.1.e.(5).

3.16.2 and 4.16.2 BASES 1.

Radiological Environmental Monitoring i

The radiological environmental monitoring program, including the land use census, is conducted to satisfy the requirements of 10CFR Part 5'J, Appendix I,Section IV.B.2 and.3.

The minimum radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathnys and for those radionuclides which lead to the highest potential radiation exposures of individuals resulting from the station operation. This monitoring program thereby supplements the radiological effluent 4

monitoring program by verifying that the measurable concentrations of 4

radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure. pathways.

The land use census is conducted annually to identify changes in use of the unrestricted area in order to recommend modifications in monitoring l

programs for evaluating individual doses from principal exposure

~

pathways. It may be conducted by dcor-to-door survey, by aerial survey, or by consulting with local agricultural or governmental l

authorities.

In order that radiological environmental monitoring stations may be relocated to reflect current conditions, the locations of stations required by Table 3.16-1 are described in a section of the Offsite Dose Assessment Manual. Revisions thereto are administered in accordance with Specification 6.15.

IELP may conduct additional environmental monitoring exclusive of the requirements of Specification's 3.16.2 and.

6.11.1.e.

3.16.3 BASES 1.

Interlaboratory Comparison Program The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in 1sP 3.16-11 l

Amendment No. 109 l

environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

3.16.4 and 4.16.4 BASES 1.

Radioactive Waste Solids This specification implements the requirements of 10 CFR Part 50.36a(a), the General Design Criterion 60 of 10 CFR Part 50 Appendix A, and of 10 CFR Part 61.56 on characteristics of low-level radioactive wastes destined for disposal by burial.

Applicable requirements on packaging and delivery of packages of radioactive material to a carrier for transport stated in 10 CFR Part 71 and on transportation of hazardous materials-in 49 CFR 171-179 are not restated in the technical specifications.

Processing waste to meet characteristics permitted under 10 CFR Part 61.56 may include solidification, preparation for deposit in a high integrity container, or any form acceptable under Part 61 for shipment to and receipt by a licensed disposal facility or licensed radioactive waste processor.

It is intended that a Contractor may perform the waste processing provided he operates according to an NRC approved Process Control Program.

3.16-12 l

Amendme-t No. 109

i DAEC-1 e.

Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Director-Nuclear Generation and to the Chairman of the Safety Committee.

f.

Review of all Reportable Events.

g.

Review of facility operations to detect potential safety hazards, h.

Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairman of the Safety Committee.

i. Review of the Plant Security Plan and implementing procedures.

i

j. Review of the Emergency Plan and implenenting procedures, k.

Review of every unplanned. release of radioactivity to the environs for which a report to the NRC is required.

j 1.

Review of changes to the Offsite Dose Assessment Manual and changes to the Process Control Program.

6.5.1.7 Authority The Operations Committee shall:

a.

Recommend to the Plant Superintendent-Nuclear written approval or disapproval of items considered under Specification 6.5.1.6 (a) through (d) above.

l i

6.5-3 l

Amendment No. 109

.~

g,,

DAEC-1 l

g.

Any other area of facility operation considered appropriate by the Safety Connittee or the President.

h.

Design change request safety evaluations at least once per 24 months.

1.

The DAEC Fire Protection Program and implementing procedures at least once per 24 months.

j. The Process Control Program and implementing procedures at least once per 24 months.

k.

The Offsite Dose Assessment M<.nual and implementing procedures at least once per 24 nionths.

~

1.

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

Performance of activities required by the QC Program for effluent and the vendors QA Program for radiological environmental monitoring.

6.5.2.9 Authority The Safety Connittee shall report to and ady'ise the President on those areas of responsibility specified in Specifications 6.5.2.7 and 6.5.2.8.

6.5.2.10 Records Records of Safety Committee activities shall be prepared, approved and distributed,as indicated below:

a.

Minutes of each Safety Committee meeting shall be prepared, approved and forwarded to the President within 14 days following each meeting.

6.5-9 Amandnwnt No.109

DAEC-1

b. Reports of reviews encompassed by Specification 6.5.2.7 above, shall be prepared, approved and forwarded to the President within 14 days following completion of the review.

't

c. Audit reports encompassed by Specification 6.5.2.8 above, shall be forwarded to the President and to the management positions responsible for the areas audited within 30 days after comoletion of the audit.

6.5.3 Other Review and Audit 6.5.3.1 Fire Protection Inspection 6.5.3.1.1 An independent fire protection and loss prevention inspection and audit shall be performed annually utilizing either qualified offsite licensee personnel or an outside fire protection firm.

6.5.3.1.2 An inspection and audit by an outside qualified fire consultant shall be performed at intervals no greater then three years, i

I 6.5-10 l

Amendmer.t No. 109 e

DAEC-1 6.8 PLANT OPERATING PROCEDURES 6.8.1 Written procedures involving nuclear safety, including applicable check-off lists and instructions, covering areas listed below shall be prepared, and approved as specified in Subsection 6.8.2.

All procedures shall be implemented and maintained.

l 1.

Normal startup, operation, and shutdown of. systems and components of the facility.

2.

Refueling operation.

i

]

3.

Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alarms, suspected primary system leaks, and abnormal reactivity f

changes.

4.

Emergency and off-normal condition procedures.

l 5.

Preventive and corrective maintenance operations which could have l

an effect on the nuclear safety of the facility.

i 6.. Surveillance and testing requirements of equipment that could have

-an effect on the nuclear safety of the facility.

7.

Procedures' required by.the Emergency Plan.

6.8-1 Amendment No.109 1

DAEC-1 8.

Procedures required by the plant Security Plan.

- 9.

Operation of radioactive waste systems.

i

10. Fire Protection Program implementation.

11. A preventive maintenance and periodic visual examination program.to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient to as low as practical levels. This program shall also include provisions for performance of periodic systems leak tests of each system no less frequently than at refueling cycle intervals, t

i

12. Program' to ensure the capability to accurately determine the

~

airborne iodine concentration in vital areas under accident conditions, including training of personnel, procedures for monitoring and provisions for maintenance of sampling and analysis equipment.

13. Offsite Dose Assessment Manual.

l

14. Process Control Plan.

15. Quality Control Program for effluents.

ll 6.8.2 Procedures described in 6.8.1 above, and changes thereto, shall.be l_

reviewed by the Operations _ Committee and -approved by the Plant Superintendent-Nuclear prior to implementation, except as provided in 6.8.3 below.

6.8.3 Temporary minor changes to procedures described in 6.8.1 above which do-not' change the intent of the original procedure may be made udth the l

concurrence of two members of the plant management staff, at least one-

[

of whom shall hold a senior operator license. Such changes shall be documented and 'promptly reviewed by the Operations Committee and by the I

Plant Superintendent-Nuclear.

Subsequent. incorporation, 'if. necessar'y, as a permanent change, shall beAin accord with 6.8.2 above.

l l

' Amendment No.109 6.8-2

DAEC-1

'l i

10. Records of radioactive effluent monitor setpoints and setpoint determinations.

6.10.2 The following records shall be retained for the duration of the Facility Operating License.

1.

Record and drawing changes reflecting facility design modifications made to systems and eculpment described in the Final Safety Analysis Report.

2.

Records of new and irradiated fuel inventory, fuel transfers an'd assembly burnup histories.

3.

Records of f acility radiation and contamination surveys, j

4.

Records of radiation exposure for.all individuals entering radiation control areas.

5.

Records of gaseous and liquid radioactive material released to the 1

environment.

6.

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

6.10-2 Anendment No. 109

DAEC-1

'7.

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

8.

Records of in-service inspections performed pursuant to these Technical Specifications.

9.

Records of Quality Assurance activities required by the QA Manual.

10.

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

4 11.

Records oT meetings of the Operations Connittee and the Safety Committee.

12.

Records for Environmental Qualification which are covered under the provi; ions of paragraph 6.13.

13.

Records of the service lives of all hydraulic and mechanical snubbers listed on Tables 4.6-3, 4.6-4 and 4.6-5 including the date at which the service life connences and associated installation and maintenance records.

14.

Records of results of. analyses required by the radiological environmental monitoring program, e

6.10-3

. Amendment No. 109

DAEC-1 I

c.

Monthly Operating Report

' Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the NRC to arrive no later than the 15th of each month following the calendar month covered by the report.

d.

Other Reports Table 6.11-1 lists some of the routine reports required by 10 CFR Parts 20, 40, 50 and 70, including those listed in Specification 6.11.1.

e.

Annual Safety / Relief Valve Challenge A report docunenting safety / relief valve challenges shall be submitted within 60 days of January 1 each year.

f.

Semiannual Radioactive Material Release Report (1) A report of radioactive materials released from the Station shall be submitted'to the NRC within 60 days after January 1 and July 1 of each year. Each report shall include the information specified i

in item (2) below covering the preceeding six months.

(2) A Semiannual Radioactive Material Release Report shall include a summary by calendar quarter of the quantities of radioactive liquid i

.and gaseous effluents and radioactive solid waste released from the Station. The data on radioactive liquid and gaseous effluents should be reported in the format in Tables 6.ll-3a and 6.11-3b.

l The data on radioactive solid waste should-include:

1.

classification of the waste (per 10 CFR part 61).

2.

total volume shipped 3.

total radioactive 'naterial shipped (curies) 4.-

identify of principal radionuclides 5.

solidification agent 6.

physical-description of the waste (3) A summary description of any changes to the PCP or ODAM.

(4) A sumary of meteorological data collected during the year will be submitted in the semi-annual report fo' lowing January 1.

Alternatively, sumary meteorological data may be retained by Iowa Electric Light and Power Company and made available to the NRC upon request.

6.11 Amend'"ent No.109

DAEC-1 g.

Annual Radiological Environmental Report An annual report of radiological environmental surveillance activities required by Specification 3.16.2 shall be submitted to the NRC before May 1 of the following year.

Each report shall include the Tollowing information:

(1) A summary description of the radiological environmental monitoring program required by Specification 3.16.2.

(2) A map and a table of distances and directions of locations of sampling stations required in Table 3.16-1.

(3) A sumnary of the land use census required in Specification 4.16.2.2.

(4) Results of analyses of sanples required by the radiological environmental monitoring program, Table 3.16-1.

In the event some results are not available, the reasons shall be explained in the report.

In the event the missing results are obtained, they shall be submitted in a supplementary report as soon as is reasonable.

(5) An assessment of radiation doses to a member of the public likely" to be the.most exposed due to radioactive liquid and gaseous effluents released from DAEC during the year. The assessment shall 7

l be performed as described in the ODAM.

(6) Deleted 6.11-4 Amendme'nt No.109 e

DAEC-1 g.

(Continued)

(7) -Results of participation in the Interlaboratory Comparison Program.

(8)

Deviation from environmental sampling schedule.

(9) A report of all analyses in which the LLD, required by Table 3.16-2, was not achieveo.

J (10) A report of any changes in sample locations.

6.11.2 Deleted 6.11-5 Amendment No. 109 l

DAEC-1 i

4 6.11.3 UNIQUE REPORTING REQUIREMENTS Special reports shall be submitted to the Director of Inspection and Enforcement Regional Office within the time period specified for each report. These reports shall be t

sumitted covering'the activities identified below pursuant to the requirements of the applicable reference specification.

a.

Reactor vessel base, weld and heat affected zone metal test specimens (Specification 4.6.A.2).

b.

I-131 dose equivalent exceeding 50% of equilibrium value i

(Specification 4.6.B.1.h).

c.

Inservice inspection (Specification 4.6.G).

d.

Reactor Containment Integrated Leakage Rate Test (Specification 4.7. A.2.f).

~

e.

deleted f.

Fire Protection Systems (Specifications 3.13.A.3, 3.13.B.2, 3.13.B.3, 3.13.C.3, and 3.13.D.3).

9-DELETED 6.11-6 l

Amendment No.109 W

l

DAEC-1 b.

Radioactive Liquid or Gaseous Effludnt - calculated dose exceeding specified limit (Specifications 3.14.3, 3.15.3 and 3.15.4).

1.

Off-Gas System (A0G) inoperable (Specification 3.15.5).

j. Measured levels of radioactivity in an environmental sampling

~

medium determined to exceed the reporting level values of Table 3.16-3 when averaged over any calendar quarter sampling period, (specification 3.16.2.b).

k.

Annual dose to a member of the public determined to exceed 40 CFR Part 190 dose limit, (specification 3.16.1.b).

1.

Radioactive liquid waste released without treatment when activity concentration exceeds 0.01 pCi/ml, (specification 3.14.4.a).

1

~6.11-7 Amendment No.109 l

1 l

DAEC-1

')

TABLE 6-11-1 REPORTING

SUMMARY

- ROUTINE REPORTS 1

Requirement Report Timing of Submittal TS Annual Exposure Within 60 days after January 1.

20.407 Personnel Exposure Within first quarter of and Monitoring each calendar. year.

~

%20.408 Personnel Exposure Within 30 days after the on Termination of exposure of the individual Employment or Work has been determined or 90 days after date of 3

{

termination of employment

~

or work assignment, whichever,is earlier.

540.64(a)

Transfer of Source Promptly upon transfer.

Material 40.64(a)

Receipt of Source Within 10 days after Material material is received.

40.64(b)

Source Material Within 30 days after Inventory September 30 of each year.

1

}

t i

tr 6.11-8 l

knendment No.109

DAEC-1 TABLE 6-11-1 (cont)

REPORTING

SUMMARY

- ROUTINE REPORTS Requirement

_ Report Timing of Submittal 50.59(b)

Changes, Tests, Within 60 days cfter i

and Experiments January 1.

70.53 Special Nuclear Within 30 days after March Material Status 31 and September 30 of each year.

70.54 Transfer of Special Promptly upon transfer Nuclear Material 70.54 Receipt of Special Within 10 days after Nuclear Material material is received Appendix G Fracture Toughness On an individual-case basis to 10 CFR at least 3 years prior to Part 50 the date when the predicted fracture toughness levels-will no longer satisfy section V.B. of Appendix G to 10 CFR Part 50.

Appendix H Reactor Vessel Completion of tests after i

to 10 CFR Material Surveillance each capsule withdrawal.

Part 50 i

Appendix J Reactor Containment Approximately 3 months to 10 CFR Building Integrated following conduct of test.

+

Part 50 Leak Rate Test I echnical Specifications T

I i

~'

I Amendmint No. 109

4 TABLE 6.11-3a p

SEMIANNUAL RADI0 ACTIVE MATERIAL RELEASE REPORT (YEAR)

LIQUID EFFLUENTS Nuclides Released Unit Quarter Quarter strontium-89 Ci E

E strontium-90 Ci E

E cesium-134 Ci E

E cesium-137 Ci E

E iodine-131 Ci E

E cobalt-58 Ci E

E cobalt-60 Ci E

E iron-55 Ci E

E I

iron-59 Ci E

E zinc-65 Ci E

E manganese-54 Ci E

E chromium-51 Ci E

E zirconium-niobium-95 Ci E

E molybdenum-99' Ci E

E technetium-99m Ci E

E barium-l anth taum-140 Ci E

E cerium-141 Ci E

E Other (speci?y)

Ci E

E Ci E

E Ci E

E Ci E

E Ci E

E Total for period (above)

Ci E

E xenon-133 Ci E

E' xenon-135 Ci E

E 6.11-10 l

l Amendmant No.109

TABLE 6.11-3b SEMIANNUAL RADI0 ACTIVE MATERIAL RELEASE REPORT (YEAR)

GASE0US EFFLUENTS 4

1 Nuclides Released Unit Quarter Quarter

}

1.

Fission gases krypton-85 Ci E

E l

krypton-85m Ci E

E i

krypton-87 Ci E

E krypton-88 Ci E

E xenon-133 Ci E

E xenon-135 Ci E

E xenon-135m.

C1 E

E I

xenon-138 Ci E

E Others (specify)

Ci E

E Ci E

E Ci E

E E

Total for period Ci E

4 2.

Iodines iodine-131 Ci E

E iodine-133 Ci E

E i

~

iodine-135 Ci E

E Total for period Ci E

E l

3. Farticulates i

strontium-89 Ci E

E strontium-90 Ci E-E 4

cesium-134 Ci E

E cesium-137 Ci E

E barium-lanthanum-140 Ci E

r 7

Others (specify)

Ci E

E I

Ci E

E Ci E

E I

6.11-11 Amandment No. 109 l

. ~.. -.. - -.

DAEC-1

.i

-6.14 0FFSITE DOSE ASSESSMENT MANUAL (00AM) 3 i

6.14.1' Changes to the ODAM may be made by IELP provided:

4 1.

Change (s) shall be submitted to the Commission by inclusion in the next Semiannual Radioactive Material Release Report after the i

change (s) was made effective and shall contain:

a.

sufficiently detailed information to support the rationale for i

the change.

Information submitted should consist of a package of those pages of the ODAM to be changed with each page numbered and provided with an approval and date, together with appropriate' bases or evaluations justifying the change (s);

b.

a determination that the change (s) will not reduce the reliability of dose calculations or setpoint determinations to i

facilitate or assess compliance with Specifications; and c.

documentation of the fact that the change has been reviewed by the Operations Committee.

]

l 1

i 6.14-1 l,

Amendment No.109

}

{

.. =

DAEC-1 1

6.14.1 (continued) 1 2.

Change (s) to radiological environmental monitoring program required sampling station locations, 00AM Table 5-1, shall be submitted to the Commission by inclusion in the next Annual Radiological Environmental Report.

i 3.

Changes shall become effective as reviewed by the Operations Comittee and approval by the Plant Superintendent-Nuclear.

i

}

}

l I'

d W

6.14-2 i

Amendment No.109

. -. -. ~

l DAEC-1 6.15 PROCESS CONTROL PROGRAM (PCP).

6.15.1 IELP may change the Process Control Program provided:

1.

Change (s) shall be submitted to the Comission by inclusion in the next Semiannual Radioactive Material Release Report after the change (s) were made effective and shall contain:

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.

c.

documentation of the fact that the change has been reviewed by the Operations Committee.

2.

Change (s) shall be come effective as reviewed by the Operations Comittee and approval by the Plant Superintendent-Nuclear.

6.15-1 Amendment No. 109

-