ML20198H165

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Proposed Tech Specs Eliminating Surveillance Requirements for Trisodium Phosphate Dodecahydrate Aggregation & Diesel Fire Pump Battery Cell Plates Visual Insp
ML20198H165
Person / Time
Site: Waterford Entergy icon.png
Issue date: 05/23/1986
From:
LOUISIANA POWER & LIGHT CO.
To:
Shared Package
ML20198H155 List:
References
GL-85-19, NUDOCS 8605300143
Download: ML20198H165 (33)
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Text

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EMERGENCY CORE COOLING SYSTEMS

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3/4.5.2 ECCS SUBSYSTEMS - T GREATER THAN OR EQUAL TO 350*F avs LIMITING CONDITION FOR OPERATION 3.5.2 Two independent emergency core cooling system (ECCS) subsy; tee;. si.all be OPERA 8LE with each subsystem comprised of:

a.

One OPERABLE high pressure safety injection pump, b.

One OPERABLE low pressure safety injection pump, and c.

An independent OPERABLE flow path capable of taking suction from the refueling water storage pool on a safety injection actuation signal and automatically transferring suction to the safety injection system sump on a recirculation actuation signal.

APPLICABILITY:

MODES 1, 2, and 3*.

ACTION:

(,'x a.

With one ECCS subsystem inoperable, restore the inoperable subsystem

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to OPERABLE status 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in 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 />.

b.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days des-cribing the circumstances of the actuation and the total accumulated actuation cycles to date.

The current value of the usage factor for each affected safety injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

  • With pressurizer pressure greater than or equal to 1750 psia.

8605300143 860523 PDR ADOCK 05000302 PDR p

WATERFORD - UNIT 3 3/4 5-3

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a hA EMERGENCY CORE COOLING SYSTENS qv SURVEILLANCE REQUIREMENTS 1

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4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

q a.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the following valves are in the indicated positions with the valves key-locked shut:

Valve Number Valve Function Valve Position a

a.

2SI-V1556 a.

Hot Leg Injection a.

SHUT (SI-506A) b.

2SI-V1557 b.

Hot Leg Injection b.

SHUT (SI-502A)

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c.

2SI-V1558 c.

Hot Leg Injection c.

SHUT (SI-5028) d.

2SI-V1559 d.

Hot Leg Injection d.

SHUT l

(SI-5068) b.

At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

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c.

By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the safety injection system sump and cause restriction of the pump suctior.s during LOCA conditions.

This visual inspection shall be performed:

1.

For all accessible areas of the containment prior to establishing CONTAllWENT INTEGRITY, and

2.

Of the areas affected within containment'at the completion of containment entry when CONTAINMENT INTEGRITY is established.

1 d.

At least once per 18 months by:

i 1.

Verifying automatic isolation and interlock action of the shutdown cooling system from the Reactor Coolant System when the Reactor Coolant System pressure (actual or simulated) is 700 1 20 psia.

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EMERGENCY CORE COOLING SYSTEMS l

SURVEILLANCE REQUIREMENTS (Continued) l 2.

A visual inspection of the safety injection system sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion.

3.

Verifying that a minimum total of 97.5 cubic feet of solid granular trisodium phosphate dodecahydrate (TSP) is contained within the TSP storage baskets.

4.

Verifying that when a representative sample of 410.01 grams of TSP from a TSP storage basket is submerged, without agitation, in 410.1 liters of 120 + 10 F water borated within RWSP boron concentration limits, the pH of the mixed solution is raised to greater than or equal to 7 within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

5.

A visual inspection of the TSP storage baskets for evidence of aggregation and the mechanical dispersal of any aggregations found.

At least once per 18 months, during shutdown, by:

e.

Verifying that each automatic valve in the flow path actuates

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1.

to its correct position on SIAS and RAS test signals.

2.

Verifying that each of the following pumps start automatically upon receipt of a safety injection actuation test signal:

a.

Highpressuresafetyinjectionpump.

b.

Low pressure safety injection pump.

3.

Verifying that on a recirculation actuation test signal, the low pressure safety injection pumps stop, the safety injection system sump isolation valves open.

f.

By verifying that each of the following pumps required to be OPERABLE performs as indicated on recirculation flow when tested pursuant to Specification 4.0.5:

1.

High pressure safety injection pumps develop a total head of greater than or equal to 1400 nsid for pump A, 1431 psid for pump 8 and 1329 psid for pump A/B.

2.

Low pressure safety injection pump discharge pressure greater than or equal to 177 psig.

W ERFORD - UM T 3 3/4 5-5

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NPF-38-15 ATTACHMENT B 1

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SURVEILLANCE REQUIREMENTS (Continued) 2.

A visual inspection of the safety injection system sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or corrosion.

3.

Verifying that a minimum total of 97.5 cubic feet of solid grr.;!r trisodium phosphate dodecahydrate (TSP) is contained l

within the TSP storage baskets.

4.

Verifying that when a representative sample of 410.01 grams of TSP from a TSP storage basket is submerged, without agitation, in 4 1 0.1 liters of 120 + 10 F water borated within RWSP boron concentraticn limits, the pH of the mixed solution is raised to greater than or equal to 7 within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

, g tica af_the TSP stor ka+e fr u suence or A vis@ tion and the machmi, u;;p@enal_ of any aggregations 5.

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At least once per 18 months, during shutdown, by:

1.

Verifying that each automatic valve in the flow path actuates

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to its correct position on SIAS and RAS test signals.

2.

Verifying that each of the following pumps start automatically upon receipt of a safety injection actuation test signal:

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a.

High pressure safety injection pump.

b.

Low 1 pressure safety injection pump.

3.

Verifying that on a recirculation actuation test signal, the low pressure safety injection pumps stop, the safety injection system sump isolation valves open.

I f.

By verifying that each of the following pumps required to be OPERABLE performs as indicated on recirculation flow when tested pursuant to Specification 4.0.5:

V 1.

High pressure safety injection pumps develop a total head of greater than or equal to 1400 psid for pump A,1431 psid for piimp B and 1429 psid for pump A/B.

2.

Low pressure safety injection pump discharge pressure greater than or equal to 177 psig.

WATERFORD - UNIT 3 3/4 5-5

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NPF~38-16

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DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-16 The following is a request to revise Technical Specification (TS) surveillance requirement 4.7.10.1.3.c.1 to remove the requirement for inspection of Diesel Fire Pump battery cell plates.

Existing Specification See Attachment "A".

Proposed Specification See Attachment "B".

Description TS 4.7.10.1.3 delineates the surveillance requirements for each fire pump diesel starting (12-volt) battery bank and charger.

In particular, item c.1 stipulates that the batteries, cell plates and battery racks are to be checked at least once per 18 months to ensure that there is no visual indication of physical damage or abnormal deterioration. Since the diesel fire pump batteries at Waterford 3 are housed in black opaque cases, the only way to visually inspect the cell plates is through the small fill caps on the top of the battery. This type of inspection does not represent a true indication of the condition of the cell plates since bridging of the cell plate would most likely occur at the bottom. This surveillance item was identified as an unresolved item (50-382/8533-01) during an NRC Inspection as documented in Inspection Report 85-33 dated January 29, 1986.

Therefore, the requested TS change to delete the surveillance requirement for visually inspectiah he diesel fire pump batteries cell plates would satisfy the NRC unresolved item and would conform to the existing installed equipment.

Safety Analysis The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:

1.

Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of any accident previously evaluated?

Response

No.

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This proposed change only involves an elimination of the requirement for visually inspecting diesel fire pump battery cell plates which based on the batteries' design can only be accomplished by viewing the cells through the small fill caps at the top of the batteries. Since this type of inspection does not represent a true indication of the cell plates' condition and since the remaining surveillance requirements provide adequate indication of the batteries' condition, the TS change would have no impact on the probability or consequences of any accident previously evaluated.

2.

Will operation of the facility in accordance with the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response

No.

The proposed change will satisfy an NRC unresolved item identified as 50-382/8533-01. The change represents an adnanistrative change that will result in surveillance requirement consistency with the installed equipment. The proposed change introduces no new perturbations to plant operations, therefore, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?

Response

No.

The proposed change to eliminate the surveillance requirement to visually inspect battery cell plates is based on the fact that the battery enclosure prevents such an inspection at the bottom of the cell plates, which is the only area that would reflect cell plate damage or deterioration. Additionally, deletion of the cell plate inspection requirement would not have an impact on the battery's integrity since other surveillance requirements exist to ensure its operability. Therefore, the proposed change would not result in a reduction in a margin of safety.

The Commission has provided guidance concerning the application of standards for determining whether a significant hazards consideration exists by providing certain examples (48 FR 14870) of amendments that are considered not likely to involve significant hazards considerations.

Example (i) relates to a purely administrative change to Technical Specifications: for example, a change to achieve consistency throughout the Technical Specifications, correction of an error, or a change in nomenclature.

In this case, the proposed change is similar to Example (i) in that the deletion of TS surveillance requirement 4.7.10.1.3.c.! will result in the TS conforming to the installed equipment.

I Safety and Significant llazards_ Determination Based on the above Safety Analysis, it is concluded that: (1) the proposed change does not constitute a significant hazards consideration as defined by 10 CFR 50.92; and (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement.

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NPF-38-16 ATTACIDIENT A t

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3 PLANT SYSTEMS 3/4.7.10 FIRE SUPPRESSION SYSTEMS FIRE SUPPRESSION, WATER SYSTEN LINITINGCONDIQ0NF6R_ OPERATION _

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3.7.10.3 The fire suppreccion water systes shall be OPERABLE with; a.

Two fire suppressicn pumps, each with a capacity of 2009 gas, with their distfjarge glibaed to the fire suppression fneader, i

b, Separate Water supplies, each with a minimum contained volume of 237,000 gglions (33 feet), and c.

An OPERABLE flow path capabl6 of taking suction fror.a the east fire water tank and the : vest fire water tank and transferring the water through distribution piping with OPERABLE sectionalizing contral or isolation valves to the yard hysirant curb valves, the last valve 2

ahead of tne water flow alare darice on each sprinkler 6r hose.

.stendpipe, 4nd the la6L valve sheed of -the deluge vahe on each delyg or spray system required to be.0PEPA3!.E per Specifica-

i. ions 3.7.10.2, 3.7.10.4, and 3.7.10s5.

ADPLICABILITf: At M1 tints.

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PCT 30N:

a.,

Wtte ene puinp and/or One water supply ihoperable, restore at least two pumps and/c.t water svpplies to 0FERABLE status within 7 riays or prcvide an alteroute backup pump or supply, b.

With the fire supprecsion water system otherwise inoperable, establish 2 bacLup fire suppression water system within 24 hauts.

c.

The provisions of Specification? 3.0.3 and 3.0.4 are not applicaole.

SuyVEILLANCE RE.WIREME'i'3 _

4.7.13.1.1 Th.e fire suppression water system shall be demonstrated OPERABLE:

a.

At least once per 7 days hy Verifying the contained water supply

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b.

At lecst once per 31 days Oy starting the electric motor-driven pump and ep.erating it fof at least 15 minute;,.

c.

Ac least 06ce per 31 days by verifying that each valve (manual, r9 War-operatad, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position.

WATERFORO - l) NIT 3 3/4 7-29

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i PLANT SYSTEMS 7

d SURVEILLANCERJEQUIREMENTS(Conti_nued) d.

At least om;e per 12 fnonths by 15erformance of a system flush.

At least onct per 12 months by cycling each testable valve in the

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e.

flow path through at least one qomplate cycle of full travel.

f, At least once per 18 months by perforaing a system functional test

'wbich insludes simulated automatic actuati6n of the system throughout its operating sequence, and:

1.

Verifying thR each pump develops at least 2000 gpm at a total head of 100 psid by verifying at least 3 points en the pump performance curve during performance testing.

2.

Cycling hach vclve in the flow path that is not testable during plant operation through at least one comple.te cycle of full travel, and 3,

Verifying th3t each fire suppression pump starts (sequenti311y)

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to m&'Intain the fire suppre5sion water system.9rescure greater than or equa'i to 'J6.b psig.

g.

At least once par 3 years b,v perfor. ming a flw test of the system in accordance with Chapter 5, Sectidn 11 of the Fire Protection I!andbook,14th Edition, published by the fiational Fire Protection Association.

4.7.10.1.2 Each fire pump diesel engine shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying:

3..

The diesel fuel oil day storaga tank contains at least 170 gallons of fuel, and 2

The diesel starts from ambient conditions and operates for at least 30 minutes.

b.

At least once per 92 days by verifying that a sample of diesel fuel from the fuel storage tank, obtained in accordance with ASTM-0270-75, is within the acceptable limits specified in Table 1 of ASTM 0975-77 when checked for viscosity, water and sediment.

At least once per 18 months during shutdown, by subjecting the c.

diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service.

WATERFORD - UNIT 3 3/4 7-30

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.7.10.1.3 Each fire pump diesel starting 12-volt battery bank and charger shall be demonstrated OPERABLE:

a.

At least once per 7 days by verifying that:

1.

The electrolyte level of each battery is above the plates, and 2.

The overall battery voltage is greater than or equal to 12 volts, b.

At least once per 92 days by verifying that the specific gravity is appropriate for continued service of the battery.

c.

At least once per 18 months by verifying that:

1.

The batteries, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration, and 2.

The battery-to-battery and terminal connections are clean, i

tight, free of corrosion, and coated with anticorrosion material.

WATERFORD - UNIT 3 3/4 7-31

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SURVEILLANCE REQUIREMENTS (Continued) 4.7.10.1.3 Each fire pump diesel starting 12-volt battery bank and charger shall be demonstrated OPERABLE:

a.

At least once per 7 days by verifying that:

1.

The electrolyte level of each battery is above the plates, and 2.

The overall battery voltage is greater than or equal to 12 volts.

b.

At lea.st once per 92 days by verifying that the specific gravity is appropriate for continued service of the battery.

c.

At least once per 18 months by verifying that:

1.

The batteries, "^ "' " ^- and battery racks show no visual l

indication of physical damage or abnormal deterioration, and 2.

The battery-to-baAtery and terminal connections are clean, tight, free of corrosion, and coated with anticorrosion material.

WATERFORD - UNIT 3 3/4 7-31

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DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-17 The following is a request to reduce the reporting requirements related to primary coolant specific activity levels based on the issuance of Generic Letter 85-19.

Existing Sp_cc_ification See Attachment "A".

}'roposed Specification See Attachment "E".

Description The proliosed change will reduce the reporting requirements for iodine spiking from a s'hort-term report (Special Report) to an item which is to be submitted annually when the limits of Technical Specification 3.4.7 are exceeded. The l

proposed change will also revise Technical Specification Bases Section 3/4.4.7 and Administrative Controls Section 6.9,1.4 to achieve consistency throughout the Technical Specifications.

In au effort to eliminate unnecessery reporting requirements (Generic Letter i

S5-19), the CommPssion has determin6d that the existing requirements to shut down a plant if coolant activity limits are exceeded for 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> in a 12 month period can be eliminated based on such factors as the improvement of' nuclear fuel Qcality, proper fuel ma.oagement practices and existing repor. ting requireraents. The Commission also has determined that the reporting requirerents related to primary coolant specific activity levels, specifically primary coolant iodine spikes,.can be reduced from a short-term repcrt to an item which is to be included in the utility's Annual Report.

Although this change will result in a relaxation of the short-term reporting rtquirements on primary coolant specific activity levels, it does not alter the associated surveillance requirements for sampling and analysis which assures that. excessive specific activity levels in the primary coolant will be detected in sof ticien*; time to take corrective action.

Laiety Ar.alysis 7be propssed changes described above shall be deemed to involve a signi.iitant bazards consideration if there in a positive finding in any of the followf.ng creas:

1.

Will cpiration of the facility in accord nce with this proposed change ittv51ve a significer.t increase in the probability or consequences of any tecident previously evalu4ted?

'Aesponte:

No.

NS20.N 31

. The removal of the short-term reporting requirements will not have an operational impact on Waterford 3 since the relevant information will be included in the annual report, when required. The removal of the constraint requiring a facility to shut down when iodine coolant activity limits are exceeded will not have an adverse impact on safety since the change does not reduce the requirements for taking reactor coolant samples and monitoring the iodine levels.

The effect of an accident occurring during a period of operating with the iodine coolant specific activity limit exceeded would be increased thyroid doses in the event of a release. However, the effect on thyroid doses is not related to the number of hours the facility has exceeded the 800 hour0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> limit, but on the specific activity itself.

Because the iodine coolant specific activity is directly related to the way the facility is being operated, appropriate action would be taken to reduce the coolant activity to below the required level prior to exceeding 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> in a 12 month period. Therefore, no increase in the probability or consequences of any accident previously evaluated will result from these changes.

2.

Will operation of the facility in accordance with the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response

No.

The proposed change involves the reduction of the administrative reporting requirements related to iodine coolant activity limits being exceeded for periods longer than 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> iv uty 12 month period and removes the requirement for subsequently shutting down the facility after this 800 hour0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> period has been exceeded.

Because this change does not result in a change in the way the facility will be operated nor does it alter the associated surveillance requirements for sampling and analysis, imple"aenting the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?

Response

No.

Since appropriate actions based on proper fuel management practices and existing reporting requirements would be initiated well before accumulating 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> above the iodine activity limit, the proposed change would not alter the existing margin of safety.

v-a The Commission has provided guidance concerning the application of standards for determining whether a significant hazards consideration exists by providing certain examples (48 FR 14870) of amendments that are considered not likely to involve significant hazards considerations.

Example (vii) relates to a change to make a license conform to changes in the regulations, where the license change results in very minor changes to facility operations clearly in keeping with the regulations.

In this case, the proposed change is similar to Example (vii) in that the change involves the reduction of the administrative reporting requirements related to iodine coolant activity limits in accordance with the changes in regulatory interpretations documented in Generic Letter 85-19.

Safety and Significant Hazards Determination Based on the above Safety Analysis, it is concluded that: (1) the proposed change does not constitute a significant hazards consideration as defined by 10 CFR 50.92; and (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC Final Environmental Statement.

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  • ,'s NPF-38-17 ATTACllMENT A

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REACTOR COOLANT SYSTEM 3/4.4.7 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.7 The specific activity of the primary coolant shall be limited to:

a.

Less than or equal to 1.0 microcurie / gram DOSE EQUIVALENT I-131, and b.

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

APPLICABILITY:

MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1, 2, and 3*:

a.

With the specific activity of the primary coolant greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of the line) shown on Figure 3.4-1, operation may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 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 / 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 any consecutive 6 month period, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days indicating the number of hours above this limit.

The provisions of Specification 3.0.4 are not applicable, b.

With the specific activity of the primary coolant greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STAN0BY with T less avg 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/ gram, be in at least HOT STANDBY with T less avg than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

With T,yg greater than or equal to 500*F.

WATERFORD - UNIT 3 3/4 4-24

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REACTOR COOLANT SYSTEM

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LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

(Continued)

MODES 1, 2, 3, 4, and 5:

d.

With the specific activity of the primary coolant greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries/ gram, perform the sampling and analysis require-ments of item 4 a) of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits.

A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days with a copy to the Director, Nuclear Reactor Regulation, Attention:

Chief, Core Performance Branch, and Chief, Accident Evaluation Branch, U.S. Nuclear Regulatory Commission, Washington, D.C.

20555.

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

1.

Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 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 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded, 4.

History of degassing operation, if any, starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded, and 5.

The time duration when the specific activity of the primary coolant exceeded 1.0 microcurie / gram DOSE EQUIVALENT I-131.

l SURVEILLANCE REQUIREMENTS 4.4.7 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

WATERFORD - UNIT 3 3/4 4-25

TABLE 4.4-4

.h PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE To AND ANALYSIS PROGRAM o

TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN kHICH SAMPLE FREQUENCY AND ANALYSIS REQUIRED g

AND ANALYSIS G

1.

Gross Activity Determination At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 1, 2, 3, 4 2.

Isotopic Analysis for DOSE 1 per 14 days 1

EQUIVALENT I-131 Concentration 3.

Radiochemical for E Determination 1 per 6 months

  • 1 4.

Isotopic Analysis for Iodine a)

Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 18, 2#, 3#, 4#, 5#

Including I-131, 1-133, and I-135 whenever the specific activity exceeds 1.0 pCi/ gram, DOSE t

EQUIVALENT I-131 or 100/E pCi/ gram, and

  • 4 b)

One sample between 1, 2, 3 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 % of the RATED THERMAL POWER within a 1-hour i

i period.

  • Sample to be taken after a minimum of 2 EFP0 and 20 days of POWER OPERATION have elapsed since reactor wds last subCritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.
  1. Until the specific activity of the primary coolant system is restored within its limits.

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. 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4 '.

DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY > 1.0 pCi/ GRAM DOSE EQUIVALENT I-131 WATERFORD - UNIT 3 3/4 4-27

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.': ?

REACTOR COOLANT SYSTEM BASES CHEMISTRY (Continued) the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant.

The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent.

Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.

The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3/4.4.7 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2-hour doses at the site boundary will not exceed an appro-priately 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 gpm and a concurrent loss-of-offsite electrical power.

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 Waterford Unit 3 site, such as site boundary location and meteorological conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1 microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

Operation with specific activity levels exceeding 1 microcurie / gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.4-1 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% of the unit's yearly operating time) since the act'ivity levels allowed by Figure 3.4-1 increase the 2-hour 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 microcurie / gram DOSE EQUIVALENT I-131 will allow sufficient time for Commission evaluation of the circumstances prior to reaching the 800-hour limit.

WATERFORD - UNIT 3 8 3/4 4-5

.'I I REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued)

Reducing T,yg 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.

3/4.4.8 PRESSURE / TEMPERATURE 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 Section 3.9.1.1 of the FSAR.

During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

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 stresses 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.

Consequently, 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.

WATERFORD - UNIT 3 8 3/4 4-6

ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted.

STARTUP REPORT

6. 9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an Operating License, (2) amendment to the license irvolving a planned increase in power level, (3) insta11aticn of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications.

Any corrective actions that were required to obtain satisfactory operation shall also be described.

Any additional specific details required in license conditions based on other commitments shall be included in this report.

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

If the startup report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial operation), supplementary reports shall be submitted at least every 3 months until all three events have been completed.

ANNUAL REPORTS 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year.

The initial report shall be submitted prior to March 1 of the year following initial criticality.

6. 9.1. 5 Reports required on an annual basis shall include a tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functionc* (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance

[ describe maintenance], waste processing, and refueling).

The dose assignments to various duty functions may be estimated leased on pocket dosimeter, TLO, or film badge measurements.

Small exposures totalling less than 20% of the indi-vidual total dose need not be accounted for.

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

"This tabulation supplements the requirements of S20.407 of 10 CFR Part 20.

WATERFORD - UNIT 3 6-17

s'.

    • .s
    • s e

NPF-38-17 ATTACHMENT B t_

o 4

!.'! I REACTOR COOLANT SYSTEM 3/4.4.7 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.7 The specific activity of the primary coolant shall be limited to:

a.

Less than or equal to 1.0 microcurie / gram DOSE EQUIVALENT I-131, and b.

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

APPLICABILITY:

MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1, 2, and 3*:

DELETE

~

With the specific activity of the primary coolant greater than icrocurie/ gram DOSE EQUIVALENT I-131 but within the able limit and to the left of the line) shown on e 3.4-1, operation ma tinue for up to A8 hours pro that the cumulative operatin under these stances does not exceed 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> in any consecu 2

period. With the total cumulative operating time ry coolant specific activity greater than 1.0 mic e/ gram DOS VALENT I-131 exceeding 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in onsecutive 6 month perio:,

are and submit a Special rt to the Commission pursuant to Speci ion 6.9.2 w

30 days indicating the number of hours above this t.

The provisions of Specification 3.0.4 are not applicable.

oL WI With the specific activity of the primary coolant greater than j

1.0 microcurie / gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T**9 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 />.

6 g(

With the specific activity of the primary coolant greater than j

100/E microcuries/ gram, be in at least HOT STANDBY with T"V9 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 />.

=

With T,yg greater than or equal to 500'F.

WATERFORD - UNIT 3 3/4 4-24 L

.': l REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

(Continued)

MODES 1, 2, 3, 4, and 5:

C. /.

With the specific activity of the primary coolant greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries/ gram, perform the sampling and analysis require-ments of item 4 a) of Table 4.4-4 until the specific activity of pn E

the primary coolant is restored to within its limits { A Special eport shall be prepared and submitted t.o the Commission pursu t

ecification 6.9.2 within 30 days with a copy to the Dir tor, Nucle Reactor Regulation, Attention:

Chief, Core Perf ance

Branch, Chief, Accident Evaluation Branch, U.S. N ear Ry ulatory ission, Washington, D.C.

20555.

T*

report shall contain the re ts of the specific activity an ses together with the followin nformation:

1.

Reactor power histo starting 48 ours prior to the first sample in which the 11 was

ceeded, 2.

Fuel burnup by core reg" 3.

Clean-up flow his y starting 48 h s prior to the first sample in whic e limit was exceede,

4.

History degassing operation, if any, star '

48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior the first sample in which the limit was ceeded, and 5.

e time duration when the specific activity of the pr ry coolant exceeded 1.0 microcurie / gram DOSE EQUIVALENT I-13.

SURVEILLANCE REQUIREMENTS 4.4.7 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

WATERFORD - UNIT 3 3/4 4-25

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p ::

  • REACTOR COOLANT SYSTEM BASES CHEMISTRY (Continued) the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant.

The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent.

Corrosion studies show that oceration may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.

The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3/4.4.7 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2-hour doses at the site boundary will not exceed an appro-priately 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 gpm and a concurrent loss-of-offsite electrical power.

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 Waterford Unit 3 site, such as site boundary location and meteorological conditions, were not considered in this evaluation.

DELETE he ACTION statement permitting POWER OPERATION to continue for limit time pe with the primary coolant's specific activity greater th-1 microcurie DOSE EQUIVALENT I-131, but within the allowabJe 1mit shown on Figure 3.

accomodates possible iodine spiki enomencn which may occur following chan.

in THERMAL POWER.

Oper with specific activity levels exceeding 1 microcurie DOSE EQUIVA I-131 but within the limits shown on Figure 3.4-1 must b s

ed 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% of the unit' e

operating time) since the act'ivity levels allowed by Figu 4.4-1 increa he 2-hour thyroid dose at the site boundary by a f r of up to 20 following stulated steam generator tube rupture.

The rting of cumulative operating time r 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in any 6-month c cutive period with greater than 1 microcurie DOSE EQUIVAL

-131 will allow sufficient time for Commission evaluatio the ci stances prior to reaching the 800-hour limit.

l WATERFORD - UNIT 3 8 3/4 4-5

u:

ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted.

STARTUP REPORT

6. 9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications.

Any corrective actions that were required to obtain satisfactory operation shall also be described.

Any additional specific details required in license conditions based on other commitments shall be included in this report.

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

If the startup report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial operation), supplementary reports shall be submitted at least every 3 months until all three events have been completed.

ANNUAL REPORTS 6.9.1.4 Annual reports covering the activities of the unit as described below gg for the previous calendar year shall be submitted prior to March 1 of each ATTACHED year. The initial report shall be submitted prior to March I of the year panne RMH following initial criticality.

"6.9.1.5 Reports required on an annual basis shall include a tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions * (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance

[de' scribe maintenance], waste processing, and refueling).

The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLO, or film badge measurements.

Small exposures totalling less than 20% of the indi-vidual total dose need not be accounted for.

In tha aggregate, at least 80% of i

the total whole body dose received from external sources should be assigned to specific major work functions.

"This tabulation supplements the requirements of $20.407 of 10 CFR Part 20.

WATERFORD - UNIT 3 6-17 t

f't

,.E ADMINISTRATIVE CONTR0t.$

ANNUAL REPORTS (Add the following to this section)

The results of specific activity analysis ih which the primary coolant exceeded the limits of Specification 3.4. 7 shall be submitted annually in accordance with the aforementioned time frame. The following information shall be included:

(1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radiolodine performed prior to exceeding the limit, results of analyhis while limit was exceeded and results of one anal-ysis after the radiolodine activity was reduced to less than limit.

Each result should include date and time of sanpling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radio-iodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiolodine limit.

TNSERT THE A BoVE AS T HE SEC oMb PARAG W M trJ SECT i o N 6 9.1,41 IMmebin rc ty t'RECEbi MG SECTi o M (o. 9.1. 5 l

1 3

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