ML17345B218

Revised Tech Specs Re Dose Equivalent Method of Calculating Radioiodine Activity in Primary Coolant

ML17345B218
Person / Time
Site:	Turkey Point
Issue date:	08/18/1983
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17345B217	List: L-83-451, Application to Amend Licenses DPR-31 & DPR-41,revising Tech Specs Re Dose Equivalent Method of Calculating Radioiodine Activity in Primary Coolant ML17345B218
References
NUDOCS 8308220317
Download: ML17345B218 (44)

Text

Section 0

TABLE OF CONTENTS Title

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TECHNICAL SPECIFICATIONS 1.0 1.1 1.2 1.3 1.0 1.5 1.6 1.7 1.8 1.9 1.10 1.11 1.12 1.13 1.10 1.15 1.16

l. 1'7 1.18 1.19 1.20 1.21 1.22

'.23 1.20 2.0 2.1 2.2 2.3 DEFINITIONS Safety Limits Limiting Safety System Settings Limiting Conditions for Operation Operable Containment Integrity Protective Instrumentation Logic Instrumentation Surveillance Shutdown Power Operation Refueling Operation Rated Power Thermal Power Design Power Dose Equivalent I-131 Power Tilt Interim Limits Low Power Physics Tests Engineered Safety Features Reactor Protection System Safety Related Systems and Components Per Annum Reactor Coolant System Pressure Boundary Integrity Coolant Loop E-Average Disintegration Energy SAFETY LIMITSAND LIMITINGSAFETY SYSTEM SETTINGS Safety Limit, Reactor Core Safety Limit, Reactor Coolant System Pressure Limiting Safety System Setting, Protective Instrumentation 1-1l-l l-l l-l l-l 1-2 1-2 1-3 1-3 1-0 I-0 I-0 I-0 1-0 1-5 1-5 1-6 1-6 1-6 1-6 1-6

,1-6 1-6 1-7 1-7 2.1-1 2.1-1 2.2-1 2.3-1 3.0 3.1 3.2 3.3 LIMITING CONDITIONS FOR OPERATION Reactor Coolant System Operational Components Pressure-Temperature Limits Leakage Maximum Reactor Coolant Activity Reactor Coolant Chemistry DNB Parameters Control Rod and Power Distribution Limits Control Rod Insertion Limits Misaligned Control Rod Rod Drop Time Inoperable Control Rods Control Rod Position Indication Power Distribution. Limits In-.Core Instrumentation Axial Offset Alarms Containment 83082203i 7 8308l8

'PDR ADOCN 050002SO PDR 3.0-1 3.1-1 3.1-1 3.1-2 3.1-0 3.1-5 3.1-6 3.1-7 3.2-1 3.2-1.

3.2-2 3.2-2 3.2-2 3.2-3 3.2-3 3.2-7 3.2-8 3.3-1

4k 0

\\

tf Ib'j, 4

~Fi ere 2.1-1 2.1-1a 2.1-1b 2.1-2 3.1-'1 3.1-1a 3.1-1b 3.1-1c 3.1-1d 3.1-2 3.1-2c 3.1-2d 3.2-1 3.2-1a 3.2-1b 3.2-1c 3.2-2 3.2-3 3.2-0 0.12-1 6.2-1 6.2-2 0

LIST OF FIGURES Title Reactor Core Thermal and Hydraulic Safety Limits, Three Loop Operation Reactor Core Thermal and Hydraulic Safety Limits, Three Loop Operation Reactor Core Thermal and Hydraulic Safety Limits, Three Loop Operation Reactor Core Thermal and Hydraulic Safety Limits, Two Loop Operation DOSE EQUIVALENTI-131 Primary Coolant Specific ActivityLimit Versus Percent of RATED POWER with the Primary Coolant Specific Activity>1.0 p Ci/gram Dose Equivalent I-131 Reactor Coolant System Heatup and Cooldown Pressure Limits Reactor Coolant System Heatup and Cooldown Pressure Limits Reactor Coolant System Heatup and Cooldown Pressure Limits Reactor Coolant System Heatup and Cooldown Pressure Limits Radiation Induced Increase in Transition Temperature for A302-B Steel Radiation Induced Increase in Transition Temperature for A302-B Steel Radiation Induced Increase in Transition Temperature for A302-B Steel Control Group, Insertion Limits for Unit 0, Three Loop Operation Control Group Insertion Limits for Unit 0, Two Loop Operation Control Group Insertion Limits for Unit 3, Three Loop Operation Control Group Insertion Limits for Unit 3, Two Loop Operation Required Shutdown Margin Hot Channel Factor Normalized Operating Envelope Maximum Allowable Local KW/FT Sampling Locations Offsite Organization Chart Plant Organization Chart B3.1-1 B3.1-2 B3.2-1

~

B3.2-2 Effect of Fluence and Copper Content on Shift of RTNDT for Reactor Vessel Steels Exposed to 550 F Temperature Fast Neutron Fluence (E>1 MEV) as a function of Effective Full Power Years Target Band on Indicated Flux Difference as a Function of Operating Power Level Permissible Operating Band on Indicated Flux Difference as a Function of Burnup (Typical)

0

1.10 DOSE E UIVALENT.I-131 The DOSE EQUIVALENTI-131 shall be that concentration of 1-131'(microcurie/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, I-133, 1-130 and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-10800, "Calculation of Distance Factors for Power and Test Reactor Sites", or in NRC Regulatory Guide 1.109, Rev.

1 October, 1977.

'1.15 POWER TILT The power tilt is the ratio of the maximum to average of the upper out-of-core normalized detector currents or the lower out of-core normalized detector currents whichever is greater.

If one out-of-core detector is out of service, the remaining three detectors are to be used to compute the average.

1.16 REACTOR COOLANT PUMPS The reactor shall not be operated with less than three reactor coolant pumps in operation.

1.17 LOW POWER PHYSICS TESTS Low power physics tests are below a nominal 5% of rated power which measure fundamental characteristics of the reactor core and related instrumentation.

1.18 ENGINEERED SAFETY FEATURES Features such as containment, emergency core cooling; and containment atmospheric cleanup systems for mitigating the consequences of postulated accidents.

1.19 REACTOR PROTECTION SYSTEM

'Systems provided to act, if'needed,

.to avoid exceeding a safety limit in anticipated transients and to activate appropriate engineered safety features as necessary.

1-5

0 0

1.20 SAFETY RELATED SYSTEMS AND COMPONENTS Those plant features necessary to assure the integrity of the reactor coolant pressure boundary, the capability to shutdown the reactor and maintain it in a safe shutdown condition, or the capability to prevent or mitigate the consequences of accidents which could result in off-site exposures comparable to the, guideline exposures of 10 CFR 100.

1.21 PER ANNUM During each calendar year.

1.22 REACTOR COOLANT SYSTEM PRESSURE BOUNDARYINTEGRITY For purposes of low temperature RCS overpressure protection, the RCS will have pressure boundary integrity UNLESS the RCS is open to containment and the minimum area of the RCS opening is greater than 2.20 square inches.

1.23 COOLANT LOOP Each of the following is defined as being a'Coolant Loop:

1.

Reactor Coolant Loop A and its associated reactor:coolant pump,and steam generator with secondary side level, greater than or equal to 10%.

2.

Reactor Coolant Loop,B and its associated reactor coolant pump and steam generator with secondary side level greater than or equal to 10%.

3.

Reactor Coolant Loop C and its.associated reactor coolant pump and steam generator with secondary side level greater than or equal to 10%.

0.

Residual Heat Removal Loop A and its associated residual heat removal pump and heat-exchanger.

5.

Residual Heat Removal Loop B and its associated residual heat removal pump and heat exchanger.

1-6

0

1.20 E - AVERAGE DISINTEGRATIONENERGY E shall be the average (weighted in proportion to the.concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines,'ith, half lives greater than 30 minutes, making up at least 95% of the total noniodine activity in the coolant.

1-7

0

e.

After shutdown, corrective action shall be taken before operation

's resumed.

f.

Above 2'f rated power, two leak detection systems of different principles shall be operable one of which is sensitive to

.radioactivity.

g.

Reactor Coolant System leakage shall be limited to 1 gpm total primary-to-secondary leakage through all steam generators not isolated from the Reactor Coolant System and 500 gallons per day through. any one steam generator not isolated from the Reactor Coolant System.

0.

'MAXIMUMREACTOR COOLANT ACTIVITY The specific activity of the primary coolant shall be limited to:

a.

Less than

.or equal to 1.0 microcurie per gram DOSE EQUIVALENTI-131, and b.

Less than or equal to 100/E microcuries per gram.

With the above limits being exceeded, the following actions shall be taken:

1.

When the reactor is critical or average reactor coolant

. temperature is reater than 500 F:

a.

With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT. I-131 but within the allowable limit (below and to the left of the line) shown on Figure 3.1-1, operation may continue for up to 08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> provided that the cumulative operating time under these circumstances does not exceed 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> in any consecutive 12 month period.

With the total cumulative operating time at a primary coolant specific activity greater.than 1.0 microcurie per gram DOSE EQUIVALENT I-131 exceeding 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in 3.1-5

0

any consecutive 6 month period, prepare and submit a Special Report to

.the Commission pursuant to Specification 6.9.3 within 30 days indicating the number of hours above this limit.

b.

With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 for more than 08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.1-1, be in a SHUTDOWN condition with average reactor coolant temperature less than 500 F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c.

With the 'specific activity of the primary coolant greater than 100/ E microcuries per gram, be in,a 'SHUTDOWN condition with average

~ reactor coolant 'temperature less than 500 F

within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

2.

For all modes of o eration a.

With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE. EQUIVALENT I-131 or greater than 100/ E microcuries per gram, perform the sampling and analysis requirements of item l.h.l of Table 0.1-2 until the specific activity of the primary coolant is restored to within its limits.

A REPORTABLE OCCURRENCE shall be prepared and submitted

.to the Commission pursuant to Specification 6.9.2.b.

This report shall contain the results of the specific activity analyses together with the following information:

1.

Reactor power history starting 08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to the first sample in which the limit was exceeded, 2.

Fuel burnup by core region, 3.

Clean-up flow history starting

08. hours prior to the first sample in which the limit was exceeded, 3.1-5a

0.

History of de-gassing operations, if any, starting 08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to the first sample in which the limit was exceeded, and.

5.

The time duration when the specific activity of the C

primary coolant exceeded 1.0 microcurie per gram DOSE EQUIVALENTI-131.

3.1-5b

41 Ot

0 E

Q 250 f

R 0 200

+ >so 8

I so 0

I

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LE ACCEPTAB OP ERATlON UNACCEPTABLE OPERATlON 60 70 80 90 100 PERCENT OF RATED POWER FIGURE 3.1-1 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED'POMER with the Primary. Coolant Specific Activity >

1.0pCi/gram Dose Equivalent X-131.

4!

0

TABLE 0.1-2 (Sheet 1 of 3 MINIMUMFRE UENCIES FOR E UIPMENT AND SAMPLING TESTS Check FrecrFuenc Max. Time Between Tests Days 1.

Reactor Coolant Samples a)

Radiochem. (Tl/2 >30 Min) b)

Cl and O2 and F c) Tritium Activity d)

Gross 9, y Activity(p Ci/cc) e)

Boron Concentration f)

E Determination g) Isotopic Analysis for DOSE EQUIVALENTI-131 Concentration, h) Isotopic Analysis for Iodine including I-l'31, 1-133 and I-135 Monthly 5/Week Weekly 5/Week 2/Week Semi-annually Biweekly 05 3

10 3

5 30 Weeks 18 1)

Once per 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> whenever the specific activity exceeds 1'.0 p Ci/gm DOSE EQUIVALENTI-131 or 100/E pCi/gm 2)

One sample be-NA tween 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a thermal power change exceeding 15 percent'of the RATED POWER within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.

2.

Refueling Water Storage Tank Water Sample 3.

Boric Acid Tank 0.

Boron Injection Tank+

~

5.

Control Rods Boron Concentration Boron Concentration Boron Concentration

'Rod drop times of all full length rods Weekly t 2/Week Monthly i For all rods at least once per 18 months and following each

,'removal of the reactor vessel head.

For specifically affected

'individual rods following maintenance on or modification of the control rod drive system which could affect the drop time of those specific rods.

10 See. reference (11) on. Page B3.0-2 Until the specific activity of the primary coolant system is restored within its 'limits.

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.TABLE 0.1-2 (Sheet 2 of 3 MINIMUMFRE UENCIES FOR E UIPMENT AND SAMPLING TESTS Check FrecrFuenc Max. Time Between Tests Days 5.

Control Rods (cont'd)

Partial movement of full length rods Biweekly while critical 20 6.

Pressurizer Safety Valves Set Point Each refueling shutdown NA 7.

Main Steam Safety Valves Set Point Each refueling shutdown 3.

Containment Isolation Trip Functioning Each refueling shutdown 9.

Refueling System Interlocks Functioning Prior to each refueling

10. Accumulator Boron Concentration At least once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each, solution volume increase of >1%

of tank volume.

11. Reactor Coolant System Leakage

12. Diesel Fuel Supply

13. Spent Fuel Pit

10. Secondary Coolant Evaluate Fuel inventory Boron Concentration I-131 Concentration Daily Weekly Prior to refueling Weekly+ g 10 NA 10

15. Vent Gas and Particulates I-131 and Particulate Activity Weekly+

10

16. Fire Protection Pump and Power Supply Operable Monthly

17. Turbine Stop and Control Valves, Reheater Stop and Intercept Valves Closure

,Monthly+++

05

13. LP Turbine Rotor Inspector V, MT, PT (w/o rotor disassembly)

Every 5 years 6 years

19. Spent Fuel Cask Crane Interlocks Functioning Within 7 days 7 days when crane is being used to maneuver spent fuel cask.

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B3.1 BASES FOR LIMITINGCONDITIONS FOR OPERATION REACTOR COOLANT SYSTEM 1.

0 erational Com onents Tge specification requires that a sufficient number of reactor coolant pumps be operating to provide coastdown core cooling in the event that a loss of flow occurs.

The flow provided will keep DNBR well above 1.30.

When. the boron concentration of the Reactor Coolant System is to be reduced, the process must be uniform to prevent sudden reactivity changes in'he reactor.

Mixing of the reactor coolant willbe sufficient to maintain a uniform boron concentration if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place.

The residual heat removal pump will circulate the reactor coolant system volume in approximately one half hour.

Each of the pressurizer safety valves is designed to relieve 283,300 lbs. per hour of saturated steam at the valve set point.

Below 350 F and 050 psig in the Reactor Coolant System, the Residual Heat Removal System can remove decay heat and thereby control system temperature and pressure.

If no residual heat were removed by any of the means available,.the amount of steam which could be generated at safety valve lifting pressure would be less than the capacity of a single valve.

Also, two safety valves have capacity greater than the maximum surge rate resulting from comples loss of load. (2)

The 50 F limit on maximum differential between steam generator secondary water temperature and reactor coolant temperature assures that the pressure transient caused by starting a reactor coolant pump when cold leg temperature is 275 F can be relieved by operation of one Power Operated Relief Valve (PORV). The 50 F limit includes instrument error.

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 power operation with one reactor coolant loop not in operation, this specification requires that the plant be in at least Hot Shutdown within 1

hour.

B3.1-1 Amendment Nos. 87 R 81

In Hot Shutdown, a single reactor coolant loop provides sufficient heat removal capability for removing decay

heat, however,,

single failure considerations require that two loops be operable.

In Cold Shutdown, a single. reactor coolant loop or RHR coolant loop provides-sufficient heat removal capability for removing decay heat, but single failure considerations require that at 'least two loops be operable.

Thus, if the reactor coolant loops are not operable, this specification requires two RHR loops to be

'operable.

The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.

The reactivity change rate associated'ith boron reduction will, therefore, be within the capability of operator recognition and control.

The requirement that at least one residual heat removal (RHR) loop be in operati'on during Refueling Shutdown ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in.the reactor pressure vessel below 160 F as required during Refueling Shutdown and (2) sufficient-coolant circulation is maintained through the

reactor, core to minimize the effect of a boron dilution stratification.

The requirement.to'have two RHR loops operable when there is less than 23 feet of water above the core ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability.

With the reactor vessel head removed and 23 feet of water above the core, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the'core.

2.

Pressure/Tem erature Limits All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads. are introduced by normal load transients, reactor trips and startup and shutdown operations.

The various categories of load cycles used for design purposes are provided in B3.1-la Amendment Nos. 87.R 81

41

Section 0.1.5 of the

. FSAR.

During startup and

shutdown, the rates of temperature and pressure chang'es are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for prevention of brittle fracture.

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall.

These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.

Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) represents a

lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.

The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location, The thermal gradients established during heatup produce tensile streses at the outer wall of the vessel.

These stresses are additive to the pressure induced tensile stresses which are already present.

The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined.

Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.

The heatup limit curves are composite curves prepared by determining the most conservative

case, with either the inside or outside wall controlling, for any heatup rate up to 100 F per hour.

The cooldown limit curves are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The heatup and coo}down curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of the service period.

B3.1-2 April 20, 1977

4i I~

i

The reactor vessel materials have been tested to determine their initial RTNDT.

Adjusted reference temperatures, based upon the fluence and copper content of the material in question, are then determined.

The heatup and cooldown limit curves include the shift in RTNDT at the end of the service period shown on the heatup and cooldown curves.

The actual shift in.NDTT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-73, reactor vessel material irradiation 'surveillance specimens installed near the inside wall of the reactor vessel in the core area.

Since the neutron spectra at the irradiation samples has a definite relationship to the spectra at the vessel inside radius, the measured transition shift for a sample can be related with confidence to the adjacent section of the reactor vessel; The heatup and cooldown curves must be recalculated when the BRTNDT determined from the surveillance capsule is different from the calculated hRTNDT for the equivalent capsule radiation exposure.

The pressure-temperature limit lines. shown for reactor criticality and for inservice 'leak and hydrostatic testing have been provided to assure compliance with,the minimum temperature requirements of.Appendix G to 10 CFR 50.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 0.2-1 to assure compliance, with the requirements of Appendix'H to 10 CFR Part 50.

The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

B3.1-3 April 20, 1977

0

3.

L~eaka e

Any leakage from the reactor coolant

system, or 'from any other system containing:potentially radioactive

material, is considered to be of major importance as it may indicate a condition is developing that would lead to gross leakage.

Gross leakage must be prevented to minimize any remote possibility of release of activity to and from the site.

Leakage prevention first of all protects the public and also it prevents potential contamination of the equipment.

Prompt maintenance and repair leads to improved reliability, which is an overall operating objective.

Thus any indication of leakage; for example:

unbalanced water inventories, radiation monitor reading increases, boric acid crystal

deposits, insulation dampness; shall be considered to be the result of a leak and shall require immediate attention with prompt evaluation required.

Action shall be prompt as it is possible that a small leak may propagate and become a major leak.

The fact that a leak of 5 gpm, at the maximum allowed reactor coolant activity, released as airborne material without holdup or cleanup, would not exceed 10 CFR 20 limits shall not permit relaxation of the requirement that action be prompt and positive.

When a real or imagined leak is detected, the Plant Supervisor will immediately initiate a detailed investigation as to source and cause after first notifying the Plant Superintendent or his designated alternate.

Evaluation will be made by the Plant Superintendent, who willcall upon Production Department supervisors, such as the Regional Superintendent and the Superintendent of Generating Stations, as necessary for consultation.

This procedure is an established and proven one in operation of fossil fuel fired units when leaks develop, as it brings to bear the judgement of experienced persons.

When the leak has been identified, the plant management will determine by a safety evaluation whether operation may continue.

Leakage source (ex. valve

stem, pump shaft seal) shall be considered.

Make up capability and potential increased demand shall also be one of the evaluation factors.

B3.1-0

4l II~

I

Leakage in,the containment willbe detected by one or more of the following:

1.

The air particulate monitor 2.

The gas monitor 3.

The sump level recorder Changes in make up water requirements 5.

Visual inspection 6.

Audible detection Leakage to other systems will be detected by: activity changes (ex. within the component cooling system), water inventory changes (ex. tank levels).

Maximum Reactor Coolant Activit

'The limitations on the specific activity of the primary coolant ensure, that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary willnot exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM.

The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations.

These values are conservative in that specific site parameters of the Turkey Point Units 3 2 0 site, such as site boundary location and meteorological conditions, were not considered in this evaluation.

The statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microcuries/gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.1-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

Operation with specific activity levels exceeding 1.0 microcuries/gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.1-1 must be restricted to no more than 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> per year (approximately 10 percent of the unit's yearly operating time) since the activity levels allowed by Figure 3.1-1 increase the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture.

The reporting of cumulative operating time over 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in any*6 month consecutive period with greater than 1.0.microcuries/gram DOSE EQUIVALENT I-131 will allow sufficient time. for Commission evaluation of the circumstances prior to reaching the 800 hour0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> limit.

.B3.1-5

4l

Reducing average reactor coolant temperature to less than 500 F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.

Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena.

A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

5.

Maximum Reactor Coolant Ox en and Chloride Concentration By maintaining the reactor coolant chemistry within the limits specified, the integrity of the Reactor Coolant System is protected.

~3~

If these limits are 'exceeded, measures can be taken to correct the condition, e.g.,

replacement of ion exchange resin or adjustment of the hydrogen concentration in the volume control tank, and further because of the time dependent nature of any adverse effects arising from concentrations in excess of the limits, it is unnecessary to shutdown immediately since the condition can be corrected.

Thus the period of 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> for corrective action to restore the concentrations within the limts has been established.

If the corrective action has not been effective at the end of the 20 hour2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> period, then the reactor will be brought to the cold shutdown condition and the corrective action willcontinue.

6.

DNB Parameters Reactor Coolant Flow Measurements

~~~

Elbow taps are used in the reactor coolant system as an instrument device that indicates the status of the reactor coolant flow.

The basic function of this device is to provide information as to whether or not a reduction in flow rate has occurred.

The correlation between flow reduction and elbow tap readout has been well established by the following equation:

W hP

=

(

)2, where DPo is the referenced pressure differential with the hP()

Wo

~

corresponding references flow rate Wo, and hP is the pressure differential with B3.1-6

4i

the corresponding flow rate W. The full flow reference point is established during initial startup.

The low flow trip point is then established by extrapolating along the correlation curve.

References (1)

(2)

(3)

(o)

FSAR Table 0.1-3 FSAR Section 10.1.10 FSAR Section 0.2.S FSAR Section 0.2.9 B3.1-7 May 28, 1976

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. ATTACHMENTA SAFETY EVALUATION/

SIGNIFICANT HAZARDS EVALUATION The proposed change substitutes the existing requirements of the Turkey Point Technical Specifications using the guidelines set forth in the Standard Technical Specifications.

The bases for the Standard Technical Specifications are that these limits "ensure that the resulting 2-hour doses at the site, boundary will not exceed an appropriately small fraction of Part l00 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of l.0 GPM".

The Bases go on to state that the values are conservative in that site specific parameters were not considered.

Technical Specification 3.I.3 enforces the 1.0 GPM primary-to-secondary leakage.

Therefore, the Bases of the 'Standard Technical Specification applies

.to Turkey Point Units 3 8 4 as well; In addition, the maximum off-site radiological dose presented in the Turkey Point Units 3 8 4 FSAR (Section l4.3.5) is the result of a postulated Loss of Coolant Accident coincident with the containment purge valves fully open when the accident initiates, and close upon receipt of signal as designed.

The primary coolant, iodine activity utilized in the analysis is 30 uCi/gm Dose Equivalent.

The proposed Amendment is well under this assumed value, and therefore conservative.

The preceeding Safety Evaluation demonstrates that fhe generic analysis also bounds Turkey Points Units 3 Bc 4.

This proposed amendment compares closely to examples (ii) and (vi) as listed in,the "Examples of Amendments that are not considered likely to involve Significant Hazards Considerations",

48FR I 4870 ~416/83) as shown below:

ii)

The proposed changes require new survailances for Isotopic analysis of Iodine I-I 3 I, I-I 33 and I-I 35, and a lower allowable value for total R.C.S. specific activity, therefore these changes constitute an additional limitation, restriction, and control not presently included in the Technical Specifications.

vi)

The proposed changes allows the use of Dose Equivalent I-l3I instead of Total Iodine as required in the current Plant Technical Specifications, therefore, these changes may result in some increase to the prob'ability or consequence of a previously analyzed accident or may reduce in some way a safety margin, but the result of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan.

i1

Based on the-preceeding Safety Evaluation and'omparison to the Examples of Amendments that are not likely to involve significant hazards consideration, these proposed changes are not a significant safety hazard in that they do not:

I.

Involve a significant increase in the probability or consequences of an accident previously evaluated; or 2.

Create the possibility of a new or different kindlof accident from any accident previously evaluated; or 3.

Involve a significant reduction in a margin of safety.

4i ig E

STATE OF FLORIDA

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COUNTY OF PALM BEACH )

Robert E. Uhri being first duly sworn, deposes and says:

That he is Vice President Licensee herein; of Florida Power & Light Company, the That he has executed the foregoing document; that the statements made in this document are true and. correct to the best of his knowledge, information, and belief, and that he is authorized to execute the document on behalf of said Licensee'-.

Robert E. Uhrig Subscribed and sworn to before me this

~8 day or I 985.

z.8 NOTARY PUBLIC, in and for the County of Palm'Beach, State of Florida.

Notary Public, State of Ftoefa at Large My communion Expires October 30, -1983 My commlsslon expires:

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