ML20246F813

Amend 115 to License DPR-61,changing Tech Specs Re Various Operation Conditions & Surveillance Requirements,Including post-accident Sampling,Noble Gas Effluent,Rcs Vents & Sampling & Analysis of Plant Effluents

ML20246F813
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	04/24/1989
From:	Stolz J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20246F811	List: ML20246F813 ML20246F835 ML20246F844
References
NUDOCS 8905150110
Download: ML20246F813 (23)

Text

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l o UNITED STATES

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<j NUCLEAR REGULATORY COMMISSION WASHINGTON, D C. 20555 L

% . . . . . #o CONNECTICUT YANKEE ATOMIC POWER COMPANY DOCKET NO. 50-213 HADDAM NECK PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.115 License No. DPR-61 1.- The Nuclear Regulatory Comission (the Commission) has found that:

A. The application for amendment by' Connecticut Yankee Atomic Power Company (the licensee), dated July 1, 1988, as supplemented December 2, 1988 and March 1, 1989, complies with the standards and

. requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission'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 Commission; C. There is reasonable assurance (1)-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 common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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890515011(j 890424 PDR ADOCA 05000213 P PDC

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2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-61 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.115, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

I

3. This license amendment is effective as of the date of issuance, to be implemented within 30 days of issuance.

F THE NUCLEA REG 9LATORYCOMMISSION s

oh 4 . Stolz, Direct r ject Directorate vision of Reactor Projects I/II OfNce of Nuclear Reactor Regulation

Attachment:

g Changes to the Technical Specifications l Date of Issuance: April 24, 1989 l

t - - _ - - _ - _ - - _ - - _--- ---_ - _ _ - _ - _

i '

l ATTACHMENT TO LICENSE AMENDMENT NO.115 FACILITY OPERATING LICENSE NO. DPR-61

)

DOCKET N0. 50-213 l Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number

-and contain vertical lines indicating the areas of change.

Remove Insert 3-4q 3-49 3-4r 3-4r 3-4s 3-4s 3-4t. 3-4t  ;

i 3-4u 3-4v 3-15b 3-15b 3-15c 3-15c 3-45 3-45 3-46

-- 3-46a

-- 3-46b 3-46c

-- 3-46d 4-2b 4-2b 4-2c 4-2c 4-33 4-33 6-18 6-18 6-22 6-22 6-23

l

, EEAQTOR COOLANT SYSTEM 3.3.5 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION l

3.3.5.1 At least one Reactor Coolant System (RCS) vent path consisting )

of at least two valves in series capable of being powered from j 125-volt D.C. buses shall be OPERABLE

• and closed at each of the  ;

following locations: 4

a. Reactor Vessel Head, and

b. Pressurizer Steam Space.

APPLICABILITY: MODES'1, 2, 3 and 4. l ACTION:

a. With the pressurizer vent path inoperable, STARTUP and/or l POWER OPERATION may continue provided that: 1) the '

inoperable vent path is maintained closed with power removed from the valve actuator of all the valves 'in the

- inoperable vent path and 11) one power operated relief valve (PORV) and its associated block valve is OPERABLE; otherwise, i restore either the inoperable vent path or one PORV and its  !

associated block valve. to OPERABLE status within 30 days, or submit a special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause i

of the malfunction and the plans for restoring the path to OPERABLE status.

b. With the reactor vessel head vent path inoperable, j STARTUP and/or POWER OPERATION may continue provided the ,

inoperable vent path is maintained closed with power {

removed from the valve actuator of all the valves in the l inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days, or, submit a Special Report to the Commission pursuant to Specification 6.9.2 '

within the next 10 days outlining the cause of the l malfunction and the plan for restoring the path to

) OPERABLE status.

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} l  !

I

• Power to the valves may be removed.

3 4q Amendment No. 97, H5 j i

l ..

l SURVEILLANCE RE0VIREMENTS Each RCS vent path shall be demonstrated OPERABLE at least-once per 18 months by;

a. Verifying all manual isolation valves in each path are locked in the open position, i

b. Cycling each valve in the vent path through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING, and c.. Verifying flow through the RCS vent path during venting during COLD SHUTDOWN or REFUELING.

4 e em 9

3-4r Amendment No. 97, 115 e

___.__________._.m_ _ _ _ _ _ _ _ _ - _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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3.3* REACTOR COOLANT SYSTFM BASES l 3.3.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES I and 2 with one reactor coolant loop not in operation this specification requires that the plant remain below 65% power. With less than the required reactor coolant loops in operation, the plant shall be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The loop isolation valves are ~_. !

required to be OPERABLE in the operating loops in order to terminate the primary to secondary leak path in the event of a steam generator tube rupture. ,

In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank

• withdrawal accident. Single failure considerations require that three loops be OPERABLE. A single reactor coolant loop provides sufficient heat removal capability for decay heat if a bank withdrawal accident can be prevented (i.e., by opening the reactor trip system breakers or de-energizing the control rod drive lift coils). Single failure considerations require that two loops be SPERABLE. -

In MODE 4, two reactor coolant loops provide sufficient heat removal capability for removing decay heat even in the event of a bank withdrawal accident. Single failure considerations require that three loops be OPERABLE. A single reactor coolant or RHR loop provides sufficient heat l removal capability for decay heat if a bank withdrawal accident can be prevented, i.e., by opening the reactor trip system breakers or de-energizing the control rod drive lift coils. Single failure

< considerations require that two loops be OPERABLE.

In MODE 5 with reactor coolant loops filled, a single RHR loop provides sufficient heat removal capability for removing decay heat. A bank withdrawal accident is prevented by opening the reactor trip system breakers or de-energizing the control rod drive lift coils. Single failure considerations require that at least two RHR loops be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat. A bank withdrawal accident is prevented by opening the reactor trip system breakers or de-energizing the control rod drive lift coils. Single failure considerations and the unavailability ~ of the steam generators as a heat removing component require that at least two RHR loops be OPERABLE.

3-4s Amendment No. 97, 115

e .3.3.1 REACTOR COOLANT SYSTEM E>ASES 3.3.1 i REACTOR COOLANT LOOPS AND COOLANT CIRCULATION (continued) l 1

The operation of one Reactor Coolant Pump (RCP) or one RHR pump. provides j adequate flow to ensure-mixing, prevent stratification and produce gradual j reactivity changes during boron concentration reductions in the Reactor Coolant System. The' reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control. -

The restrictions on starting an RCP with one or more RCS cold legs less than or equal to 315 F are,provided to prevent RCS pressure transients, caused by i

energy additions from .the' Secondary Coolant System, which could exceed the -  ;

limits of Appendix.G to 10 CFR Part 50. The RCS will be protected against  !

overpressure transients and will not exceed the limits of Appendix G by I restricting starting of the RCPs to wheg the secondary water temperature of each steam generator is less than 20 F above each of the RCS cold leg temperatures.

The requirement to . maintain the boron concentration of an isolated / idled loop gftater'than or equal to the boron concentration of the operating loops ensures that no reactivity addition to the core could occur during startup of an isolated / idled loop. Verification of the boron concentration in an isolated / idled loop prior to opening the stop valves provides a reassurance l of the adequacy of the boron concentration in the isolated / idled loop.

Startup of an isolated / idled loep could inject cool water from the loop into the core. The reactivity transient resulting from this cool water injection is minimized by prohibiting isolated / idled loop startup until its temperature is within 20 F of the operating loops. Making the reactor suberitical prior to isolated loop startup prevents any power spike which could otherwise result from this cool water-induced reactivity transient.

3.3.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 293,300 lbs per hour of saturated steam at 2485 psig.

The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no. safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.

3-4t Amendment No. 97, 115

"c a.

. REACTOR COOLANT SYSTEM BASES i

3.3.2 SAFETY VALVES (continued)

During operation, all pressurizer Code safety valves must be OPERABLE to

. prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

The combined' relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss-of-load assuming no Reactor trip until the first Reactor Trip System Trip Setpoint is reached (i.e., no credit is taken for a direct Reactor trip on the loss-of-load) and _

also assuming no operation of the power-operated relief valves or steam dump valves.

Demonstration of the' safety valves' lift settings will occur only. during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

3.3.3 PRES $URIZER The limit on the water level in the pressurizer assures that the parameter is maintained within that assumed in the safety analyses. The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation.

L3 4 RELIEF VALVES Operation of the power-operated relief valves (PORVs) minimizes the j undesirable opening of the spring-loaded pressurizer Code safety valves and  ;

provides an alternate means of core cooling. Each PORV has a remotely '

operated block valve to provide a positive shutoff capability should a PORV become inoperable. One of two redundant PORV relief trains must be OPERABLE to adequately cool the core in the event that the steam generators are not available to remove core decay heat. When in automatic mode, all PORVs and block valves open automatically on high pressurizer pressure. The PORVs and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.

The OPERABILITY of two spring-loaded relief valves (SLRVs) or an RCS vent opening of greater than 7 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR gart 50 when one or morc cf the RCS cold legs are less than or equal to 315 F. Either SLRV has adequate relieving capability.to protect the RCS from overpressurization when the transient is limited to either: (1) the start of an idle RCP with the secondary g

water temperature of the steam generator less than or equal to 20 F above the RCS cold leg temperatures, or (2) the start of a charging pump (centrifugal) and .its injection into a-water-solid RCS. 1 3-4U Amendment No. 115

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.______u--_____-._---____-_--_._--

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The Maximum Allowed SLRV Setpoint for the Low Temperature Overpressure Protection System (OPS) is derived by analysis which models the performance of the OPS assuming various mass input and heat input transients. Operation with a SLRV Setpoint less than or equal to the maximum Setpoint ensures that .

Appendix G criteria will not be violated with consideration for a maximum pressure overshoot beyond the SLRV Setpoint which can occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single failure. To ensure that mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require lockout of all but one charging pump (centrifugal or metering) while in MODES 4, 5, and 6 with the reactor vessel head installgd and disallow start of an RCP if secondary temperature is more than 20 F ,

above RCS cold leg temperature. _.

3.3.5 REACTOR COOLANT SYSTEM VENTS

. Reactor Coolant Systeln. vents are provided to exhaust noncondensible gases and/or steam from the RCS that could inhibit natural circulation core cooling. The OPERABILITY of at least one RCS vent path from the reactor vessel head and the pressurizer steam space ensures the capability exists to perform this function.

The valve redundancy of the RCS vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single-failure of a vent valve, power supply or control system does not prevent isolation of the vent path.

The function, capabilities, and testing requirements of the RCS vents are consistent with the requirements of Item II.B.1 of NUREG-0737,

" Clarification of TMI Action Plan Requirements", November 1980.

Amendment No. 115 3-4v

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, , TABLE 3.9-2 (CONTINUED) .

1 ACTION 1 -

With the number of OPERABLE channels less than required by  ;

Table 3.9-2, either restore the inoperable channel (s) to  !

OPERABLE status within 30 days or be in 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 />.

ACTION 2 -

Not used ACTION 3 -

With any individual valve position indicator inoperable, obtain quench tank temperature, level and pressure information, and - .

monitor discharge pipe temperature once per shift to determine valve position. This action is not required if the PORY block valve is closed with power removed in accordance with Specification 3.3.C.(6).

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3-15c Amendment No. 42. 47, 115 i

___ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ i

I 1

- INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION '

3.23 The accident monitoring instrumentation channels shown in Table 3.23-1 shall be OPERABLE. f

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APPLICABILITY: As shown in Table 3.23-1 ACTION: As shown in Table 3.23-1 I

SURVEILLANCE RE0VIREMENTS Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION at the frequencies shown in Table 3.23-2 e

3-45 Amendment No. ,7d,115

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TABLE 3.23-1 (Continued)

ACTION STATEMENTS ACTION 1 -

With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels shown in Table 3.23-1, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least-HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 2- -

With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels i OPERABLE requirements of Table 3.23-1, restore the inoperable channel (s) to OPERABLE status within -48 hours or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 3 -

With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirements, return one channel to operable status within 7 days, or else prepare and submit Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining:

the cause of the malfunction, the plans for restoring the channel to OPERABLE status, and a preplanned alternative method for estimating stack release rates during- the interim.

ACTION 4 -

With the number OPERABLE channels less than the Total Number of Channels shown in Table 3.23-1, either restore the inoperable channel (s) to OPERABLE status within 7 days if repairs - are feasible without shutting down or prepare and submit a Special report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status and alternate methods in effect for estimating the applicable parameter in the interim.

ACTION 5 -

With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements, either restore the inoperable channel (s) to OPERABLE status

- within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if repairs are feasible without shutting down or;

a. Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status; and 3-46a Amendment No. 115

- -_-__-_-___ ____-_-_- -_-____--_-_ _ __-__-_ _ _ - _ a

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b.

L , . Restore the system to OPERABLE status at the next scheduled refueling.

ACTION 6 -

'With .the number of OPERABLE accident- monitoring instrumentation channels less than ~ the MINIMUM CHANNELS OPERABLE : requirements of Table 3.23-1, restore the -)

inoperablechannel(s to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, i or submit a Special ) Report to the Commission pursuant -to

. Specification 6.9.2 within the next 10 days outlining the cause of the malfunction, the' plans for restoring the channel (s) to OPERABLE status, and any alternate methods in affect for -estimating the applicable parameter during.

the interim.

i

< ACTION 7 -

With less than the minimum channel (s) operable, restore the inoperable channel (s) to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> o_r else establish; alternate means to determine .if significant fuel failure exists. .If still inoperable after 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining: the cause of the inoperability, the plans ~ for' restoring operability, and the alternate means established.

ACTION 8 -

With the number of channels operable less than the-Minimum Channels OPERABLE, determine the subcooling margin once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Restore'the system to OPERABLE-s ~s tatus at the next scheduled refueling.  ;

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3-46b Amendment No. 115

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, INSTRUMENTATION '

BASES 3.23 ACCIDENT MONIT@ LNG INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, Revision 3,

" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," May 1983 and NUREG-0737,

" Clarification of TMI Action Plan Requirements," November 1980.

3-46d Amendment No. 115

1 Table 4.2-1 (continued) 1 j ,

' Channels Action Minimum Freauency i

8. Variable Low Calibrate Each refueling .

Pressure trip set Check Each shift 1 point calculator

9. Rod Position Calibrate Each refueling Digital Voltmeter Check with counters Every six inches of rod motion when data logger is out of service.

10. Rod Position Test Each refueling -

Counters Check with Digital Every six inches of rod Voltmeter motion when data logger is out of service.

4

11. Steam Generator Calibrate Each refueling Level Check Each shift . ,

12. Steam Generator Calibrate .Each refueling Flow Mismatch Check Each shift

13. Charging Flow Calibrate Each refueling

~

14. Residual Heat Calibrate Each refueling Pump Flow-

15. Boric Acid Calibrate Each refueling Tank Level Check Each week

16. Refueling Water Calibrate Each refueling Storage Tank Test 90 days Level

17. Volume Control Calibrate Each refueling Tank Level Test 90 days

18.

• Blank 19 Radiation Calibrate Each refueling Mor.itoring Test Each day System

20. Boric Acid Ca11b ate Each refueling Control

21.

• Blank ,

22. Valve Temperature Test Each refueling Interlocks 4-2b Amendment No. 53, 115

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l TABLE 4.2-1 (CONTINUED)

Channel Action Minimum Freauency

23. Pump Valve Interlock Check Each refueling i

24. Reactor Coolant System Calibrate Each refueling  !

OPS Test .Each refueling j

25. Auxiliary Feedwater Flow Rate Calibrate Each refueling  ;

Check Each month

26.

• Blank 27 .' PORY Position Indication Calibrate Each refueling (Acoustic Monitor) Check Each month

28. PORV Block Valve' Calibrate Each refueling

. Position Indication

29. Safety Valve Position Calibrate Each refueling Indication (Acoustic Monitor) Check Each month 9

• Items 18, 21 and 26 of this Table are included in Table 3.23-1 (Items 1, 5, and10). -

4-2c Amendment No. 42, 115 j

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Amendment No. 56, 115 4-33 4

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ADMINISTRATIVE CONTROLS L , SEECIAL REPORTS 6.9.2 Special reports shall be submitted to t'he U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator, Region I, and one copy to the NRC Resident Inspector, within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

a. Inservice Inspection results, Specifications (4.10) and (4.12).

b. Primary Containment Leak Rate Results, Specification (4.4).

c. Reactor Vessel Material Surveillance Specimen Examination, Specification (4.10),

d. Steam Generator Tube Report Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.

e.- Post-Accident Instrumentation Operability, Specification (3.23 A and B).

f. Fire Protection Systems Operability, Specification (3.22).

g. Reactor , Coolant System Vents, Specification (3.3.5.1)

h. Radiological Effluent Reports required. by Specifications (7.1.1.2, 7.1.2.2, 7.1.2.3 and 7.1.3) 6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years: .

a. Records and logs of fac.ility operation covering the time interval at each power level,

b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety. 3

c. All reportable events.

d. Records of surveillance activities, inspect lons and calibrations required by these Technical Specifications.

6-18 Amendment No. 20.56,79,197,115

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. . ADMINISTRATIVE CONTROLS i

L - ' 6.1'7 RADIOLOGICAL EFFLUENT MONITORING AND OFFSITE DOSE CALCULATION MANUAL (REM 0DCM)

U Section I, Radiological Effluents Monitoring Manual, shall outline the sampling and analysis programs to determine the concentration of radioactive materials released offsite as well.as dose commitments to ' individuals in those exposure pathways and for those radionuclides released as a result of station operation. It shall also specify operating guidelines for radioactive waste treatment systems and report content.

- Changes to Section I shall be submitted ' to the Commission for-approval prior to implementation.

Section ' II, the Offsite- Dose Calculation Manual (0DCM), shall describe the. methodology 'and parameters to be used; in the calculation of.offsite doses.due to radioactive gaseous and liquid effluents and in.the calculations of- gaseous and liquid effluent monitoring instrumentation alarm / trip'setpoints consistent with the applicable LCO's contained. in these technical specifications.

Changes to Section II need not be submitted to the Commission for approval prior to implementation, but shall be included in the next Semi-Annual Radioactive Effluent Release Report.

6.18- ' RADIOACTIVE WASTE TREATMENT SYSTEMS Procedures for liquid and gaseous radioactive effluent discharges from the Unit shall be prepared, approved, maintained and adhered to for all operations involving offsite releases of radioactive effluents. These procedures shall specify the use of appropriate waste treatment systems utilizing the guidance provided in the REMODCM.

The solid radioactive waste treatment system shall be operated in accordance with_ the Process Control- Program to process wet radioactive wastes to meet shipping and burial ground requirements.

6.19- PJSS/Samolino and Analysis of Plant Effluents The licensee $ hall implement and maintain a program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulate in plant gaseous effluents, and containment atmosphere samples under accident conditions. This program shall include the following:

a. Training of personnel 6-22 AmendmentNo.$B,7Mf, 115

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. . ADMINISTRATIVE CONTROLS

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b. Procedures for sampling and analysis, and c.- Provisions'for maintenance of sampling and analysis equipment.

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6-23 Amendment No. 115 f

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