Text

Safety Evaluation Supporting Amends 137 & 132 to Licenses DPR-31 & DPR-41,respectively
	ML17348A534
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Site:	Turkey Point
Issue date:	08/28/1990
From:	; Office of Nuclear Reactor Regulation
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Op UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 SAFETY EYALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 137 TO FACILITY OPERATING LICENSE NO.

DPR-31 AND AMENDMENT NO. 132TO FACILITY OPERATING LICENSE NO.

DPR-41 FLORIDA POWER AND LIGHT COMPANY TURKEY POINT UNIT NOS.

3 AND 4 DOCKET NOS.

50-250 AND 50-251

1. 0 BACKGROUND Turkey Point Units 3 and 4 currently operate with custom Technical Specifications issued with the operating licenses in 1972 and 1973 and amended from time to time over the years.

In 1976 the Nuclear Regulatory Commission (NRC) issued NUREG 0452, "Standard Technical Specifications for Westinghouse Pressurized Water Reactors" and subsequently issued revisions, including Revision 4 in 1981.

In April 1984, Florida Power and Light Company (FPL) committed to review and implement, when appropriate, the philosophy and guidance of the STS in the development of upgraded plant procedures.

At that time FPL also committed to incorporate the requirements of the STS in all future proposed amendments.

Later in 1984, FPL voluntarily expanded the original commitment to include development of a fully revised and reformatted set of Turkey Point Technical Specifications based on Draft Revision 5 of NUREG-0452 within certain limita-tions.

Those limitations were that the RTS would not require hardware

changes, would reflect the current Turkey Point plant design and analytical basis, and would consider operating hardship or reasonable additions.

On September 29,

1986, FPL submitted a license amendment request with proposed revised and reformatted Technical Specifications (TS).

Issues identified in reviewing the proposal were discussed by the NRC staff and FPL representatives in a series of meetings, and it became apparent that significant additional work was needed on the 1986 proposal.

On June 9, 1988, the Technical Specifications Branch issued a modification of the TS, designated the Proof and Review version of the RTS.

The TS weve modified to incorporate additional information developed during the intermediate meetings.

Comments on the Proof and Review version were provided by FPL and by the NRC. staff.

Resolutions of issues raised by those comments were included in a Final Draft issued on March 14, 1989, and supplemented on May 12, 1989.

On June 5, 1989, FPL submitted a new, request fov amendment, superseding the 1986

proposal, and which was based on the Final Draft and certified that the final draft with minor changes accurately reflected the as-built facility.

On July 12,

1989, and Novembev 3, 1989, FPL transmitted a list of changes that will be incorporated into the next update of the Final Safety Analysis Report (FSAR) and certified that with those changes the Final Draft would accurately reflect the updated FSAR.

Minor changes and corrections identified since the amendment request on June 5, 1989, have been incorporated into the RTS to be implemented DOCK 05000250 I

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by these "amendments.

By letters dated June 21 and July 20, 1990, the licensee provided supplemental information.

This information did not change the scope of the amendments and did not alter the staff's proposed determination of no significant hazards consideration which was issued on May 15, 1990.

This Safety Evaluation documents the basis for the staff's conclusion that use of the RTS-is an acceptable and improved alternative to continued use of the current Technical Specifications (CTS) and that FPL's commitment to develop TS based on the STS has been fulfilled except for the Electrical Systems interim TS included in the RTS (see Section 2.5 below).

1. 1 Basis for Use of STS In developing the

STS, a basic format change from custom TS (such as Turkey Point TS) was the combination of Section 3, Limiting Conditions for Operations (LCOs), and Section 4, Surveillance Requirements (SRs),

as Section 3/4.

Six modes of operation were established by Table 1.2 of the STS, and each LCO was stated to be applicable for one or more of those six modes.

Nine frequency notations for SRs were established by Table 1. 1.

A Section 3/4.0 was added to prescribe a standardized sequence for TS-required shutdowns and other gener ally applicable specifications.

Model action statements were included with each LCO.

Each of the above standardizations was consistent with the Commission's regulatory requirements in 10 CFR 50.36.

The requirement for LCOs and SRs was preserved.

The requirement to shut down or to take such remedial action as provided by the TS if an LCO is not met was preserved by Section 3/4.0, Applic-ability, and the action statements associated with each LCO.

Action statements typically specify a time permitted to correct a situation prior to shutting down.

Standardized times have been selected for use in several typical situations.

Controls on shutting down have been standardized in terms of the times allowed to achieve recurring operational mode reductions.

As stated in FPL's Safety Evaluation included in its request for amendment (September 26, 1986 and as superseded June 5, 1989), the STS, which have been utilized by newly licensed plants, are recognized to be more prescriptive and contain an increased number and frequency of surveillances than the custom Turkey Point Technical Specifications.

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Since

1974, new operating

'licenses have, included Technical Specifications (TS) in a standard format as Appendix "A".

These standard format TS are based on a set of model TS called the Standard Technical Specifications.

These STS are based on the design and design basis safety analyses of a typical plant and are adjusted to account for differences between the plant's actual design and safety analyses and those of the typical plant.

The actual plant design and design basis safety analyses are described in the licensee's FSAR.

The NRC reviews the FSAR and issues a Safety Evaluation Report (SER) explaining the acceptability of the plant design and safety analyses.

Therefore, the SER provides the NRC's basis for granting the plant's operating license.

At the same time, the SER provides the NRC's basis for approving the plant's TS which are part of the license and based on the FSAR.

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This process has been repeated for about 60 plant operating licenses since 1974.

Therefore, the acceptability of the STS, as modified to account for differences associated with a particular plant, has been well established by the

staff, as reflected in about 60 SERs.

In addition, since 1974 nuclear power plants have amassed hundreds of reactor-years of experience using STS.

The STS have been adjusted over the years because experience has identified areas which needed or could benefit from improvement.

Also, changes in plant design since 1974 have resulted in changes in the "typical" plant design reflected by the STS.

Many parts of the STS have also been examined using probabilistic risk assessment (PRA) as a tool.

Except for rare instances which were corrected, PRA showed that the STS are conservative in terms of safety.

Following are two examples of how the STS are applied generically to an actual plant design.

In the first example, LCO 3.0.3 provides a schedule for shutting down the plant when the safety function capability of a system becomes inoperable.

The schedule allows time periods for taking the plant to cold shutdown.

These time periods are intended to shut down the plant as quickly and as safely as possible because of the lost safety function.

At the same time, the time periods are long enough to allow the plant to be shut down in a controlled and orderly manner.

This reduces the potential for transients that could challenge safety systems.

Also, this schedule allows the plant cooldown rate to be maintained well within the maximum limits, which minimizes the thermal stresses on the primary coolant system.

These time periods are based on the experience from thousands of plant shutdowns.

This LCO is applied to almost all plants with STS.

A second example of the direct application of STS to an actual plant is the allowance of up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to provide the licensee an opportunity to repair a

system and return it to service without shutting down the plant.

In the typical

design, many safety systems have two redundant trains of equipment, each train being individually capable of performing the safety function.

These redundant systems provide "single failure" protection--if one train fails the other train is completely capable of performing the safety function.

In STS, if one of these trains is inoperable, the plant is allowed to continue to operate without beginning a shutdown for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

During this time, "single failure" protection has been lost for the safety system;

however, the operable train still provides full capability if an accident occurs.

The risk of an accident occurring during the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is low.

The combined risk of an accident occurring within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the operable train failing is even lower.

On the other hand, a

plant shutdown, no matter how careful, involves some risk of initiating a transient or challenging safety systems.

The NRC concluded that the allowed limit of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provides the appropriate balance of safety.

This allowed limit is applied unrevised to almost all dual train safety systems for plants with STS.

Turkey Point was not licensed with STS; however, the TS were derived from analyses in the Final Safety Analysis Report.

The SER which supported issuance of the operating license for Turkey Point supported the acceptability of the plant design, design basis safety analyses, and TS.

The RTS approved by this Safety Evaluation do not change the plant design or design basis safety analyses for Turkey Point.

The RTS are the plant-specific application of the STS to Turkey Point in the same manner described above for new operating licenses.

Therefore, this safety evaluation relies on the original SER for Turkey Point as amended over the years and on the col1ective SERs for plants licensed with STS since 1974.

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RTS 3/4.1.3.2 is an example of a plant-specific application of STS to the Turkey Point plant.

Turkey Point uses an analog instead of a digital control rod position indication system that is governed by STS 3/4.1.3.2, Position Indication Systems-Operating.

In developing RTS 3/4. 1. 3. 2, parts of the STS were retained with little or no change.

Changes for significant design differences utilized the requirements set forth in the STS for other analog systems.

The applicability in this RTS for the analog system is the same as that for a digital system in the STS, MODES 1 and 2.

The action statements are identical except for the title "analog" instead of "digital" and a power reduction to 75%

instead of 50K of Rated Thermal Power as in the STS.

The reduction to 75K is consistent with the CTS.

Surveillance requirement (SR) 4. 1.3.2 is identical to the STS except for the title "analog" instead of "digital" and the addition of a parenthetical note allowing for one hour thermal soak after rod motion prior to performing the measurement.

The LCO has been expanded to apply to specified withdrawal ranges separately for shutdown banks for two sets of control banks and for group demand counters.

The STS limit of t 12 steps has been retained for the shutdown and control banks but has been reduced to ~ 2 steps for the group demand counters.

SR 4. 1.3.2.2 was added to require Channel

Checks, Channel Calibrations and Analog Channel Operational Tests.

These are typical SRs used in TS for other analog systems elsewhere in the STS.

On the basis that much of the STS has been retained without changes and, that when modified, it has been consistent with the general guidance of the

STS, and with reliance on the licensee's certification, we conclude that RTS 3/4. 1.3.2 is an acceptable TS for Turkey Point Units 3 and 4.

This example is illustrative of the bases for the staff's finding that the RTS are acceptable.

Finally, NRC programs to improve nuclear reactor safety by improved TS are ongoing.

As improvements are identified and found to be acceptable for a class of plants, incorporation of such improvements is achieved in individual plants by license amendments.

Such opportunities will continue to be available for Turkey Point.

1.2 Use of STS The development of the RTS consisted of modifications to the STS to account for design diff'erences and the constraints identified in FPL's letter of September 29, 1986.

Specifically, action times and SR frequencies in the STS were considered acceptable for comparable design features even though some CTS included more restrictive requirements.

Our evaluations that follow are grouped by characteristic differences from the STS.

The first group includes those CTS which were already in close agreement with the STS and were only modified to a limited extent.

The second group of RTS are for design features that do not differ significantly from the design features for the STS.

The third group of the the RTS are character ized by differences from the STS to reflect unique Tur key Point design features.

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fourth group are options to the current STS that have been found acceptable for a general class of plants.

During the development of the RTS, the licensee proposed alternate TS for Electr ic Power Systems and requested that full conversion to the RTS for these systems be deferred to a later amendment that would also provide for staff approval of planned facility modifications.

There-fore, the alternate TS for the Electric Power Systems have been incorporated as the fifth group, but only as interim TS until planned modifications to enhance the Electrical-Power Systems are completed.

2. 0 EVALUATION As noted above, acceptability of the RTS was determined by evaluating the acceptability of differences from the STS.

These differences fall into the following five groups.

2. 1 RTS Same as CTS Exce t for Editorial Differences Amendments to the CTS issued during the 'last several years have been in the general format of the STS.

About one-fifth of the TS in this conversion fall into this category but do not necessarily adhere to STS in all detail.

The differences are largely editorial in nature and do not affect technical require-ments.

Therefore, those CTS have been incorporated in the RTS with only minor additional editorial and format changes.

The SERs for those amendments continue to serve as the bases for the acceptability of those RTS.

These TS are the following:

RTS CTS Amendment No.

Title

3. 1.1.3 4.4.5 3/4. 7. 1. 2 3/4. 7. 1. 6 3/4. 7. 6 3.9.1 3.9. 2 3.9.3 3.

9.4'.9.5 3.9.8.1 3.9.8.2 3.9.9 3/4. 11. 1 3/4. 11. 2 3/4. 12 6.2.1

3. 1. 2. 1 4.2. 5

3. 18. 1/4. 10 3.20/4.21

3. 13. 1/4. 14. 1

3. 10.8

3. 10. 3

3. 10. 5

3. 10. 1

3. 10. 6

3. 10. 7. 1

3. 10.7. 2

3. 10. 2 3.9.1 3.9.2

4. 12 6.2. 1 115/109 119/113 124/118 118/112 116/110 114/108 114/108 114/108 114/108 114/108 114/108 114/108 114/108 103/97 103/97 102/97 129/123 Moderator Temperature Coefficient Steam Generators (SR only)

Auxiliary Feedwater System Standby Feedwater System Snubbers Boron Concentration Instrumentation Decay Time Containme'nt Building Penetrations Communications Residual Heat Removal and Coolant Circulation (High Water Level)

Residual Heat Removal and Coolant Circulation (Low Water Level)

Containment Ventilation Isolation System Liquid Effluents Gaseous Effluents Radiological Environmental Monitoring Onsite and Offsite Organization

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6. 2. 2i 6.5.2.2 6.5.2.6

6. 2. 2a 6.5.2.2 6.5.2.6 135/129 136/131 136/131 Operations Superintendent Composition of Company Nuclear Review Board (CNRB)

CNRB quorum 2.2 RTS Same as STS Exce t for Minor 0 eratin Parameter Differences About one-third of the RTS are the same as the corresponding STS except for minor adjustments to accommodate plant-specific operating parameters which reflect the existing licensing basis.

As indicated in Section

2. 1, the'taff considers that this is a sufficient basis for acceptability of those RTS provided the Turkey Point design corresponds to the design reflected by the STS.

We conclude, on the basis of our review and the licensee's certification, that there is such correspondence of design for the following RTS:

RTS Title 2.1.1

2. 1.2 3/4. 1. 1. 2 3/4. 1. 1. 4 3/4. 1. 2. 1 3/4. 1. 2. 4 3/4. 1. 2. 5 3/4. 1. 3. 1 3/4. l. 3. 4 3/4. l. 3. 5 3/4. l. 3. 6 3/4. 2. 3 3/4. 2. 4 3/4. 2. 5 3/4. 4. 1 3/4.4. 2. 1 3/4. 4. 2. 2 3/4. 4. 3 3/4. 4. 7 3/4. 4. 8 3/4. 4. 9. 1 3/4. 4. 9. 2 3/4. 4. 10 3/4. 4. 11 3/4, 6. 1. 1 3/4. 6. 1. 2 3/4. 6. 1. 3 3/4. 6. l.4 3/4. 6. 1. 5 3/4. 6. 2. 1 3/4. 6. 2. 4 3/4. 7. 1. 1 3/4. 7. l. 4 3/4. 7. 7 3/4.7.8.3 Reactor Core Reactor Coolant System Pressure Reactivity Control Systems-Shutdown Margin"Tavg less than 200 F

Reactivity Control Systems-Maximum Temperature for Criticality Boration Systems Flow Path Shutdown Borated Water Source-Shutdown Borated Water Source-Operating Movable Control Assemblies - Group Height Movable Control Assemblies Rod Drop Time Movable Control Assemblies Shutdown Rod Insertion Limit Movable Control Assemblies Control Rod Insertion Limits Nuclear Enthalpy Rise Hot Channel Factor quadrant Power Tilt Ratio DNB Parameter Reactor Coolant Loops and Coolant Circulation Safety Valves-Shutdown Safety Valves-Operating Pressurizer Chemistry Specific Activity Pressure Temperature Limits-Reactor Coolant System Pressure Temperature Limits-Pressurizer Structural Integrity Reactor Coolant System Vents Containment Integrity Primary Containment-Containment Leakage Primary Containment-Containment Air Locks Primary Containment-Internal Pressure Containment Air Temperature Containment Spray System Containment Isolation Valves Turbine Cycle-Safety Valves Turbine Cycle-Specific Activity Sealed Source Contamination Fire Suppression Systems-Fire Hose Stations

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3/4. 7. 8. 4 3/4. 7. 9 3/4. 9. 1 4.9.2, 4.9.3

4. 9.4 4.9.5 3/4.9.6

4. 9.8.2 4.9.9 3/4.9.10 3/4. 10. 1 3/4. 10. 2 3/4. 10. 3 3/4. 10. 5 3/4. 11. 3 3/4. 11. 4 5.0 6.0 Fire Suppression Systems-Fire Hydrants and Hydrant Hose Houses Fire Rated Assemblies Refueling Operations Instrumentation Decay Time Containment Building Penetrations Communications Manipulator Crane Residual Heat Removal and Coolant Circulation-Low Water Level Containment Ventilation Isolation System Water Level-Reactor Vessel Shutdown Margin Group Height, Insertion, and Power Distribution Limits Physics Tests Position Indication System-Shutdown Solid Radioactive Wastes Total Dose Design Features Administrative Controls except 6.2.1, 6.2.2i, 6.5.2.2 and 6.5.2.6

2. 3 RTS Like STS Exce t for Desi n Differences About one-third of the RTS required some modification of the STS to accommodate design differences.

This tailoring is a common step in applying STS in the development of TS for operating licenses.

Our approach for those TS and for the RTS has been to modify the STS only as necessary to accommodate the design differences which reflect the current licensing basis.

We conclude, on the basis of our review and on the licensee's certification, that the accommodation of the design differences in the following RTS is consistent with STS format and guidance.

RTS Title Desi n Difference 2.2.1 Reactor Trip System Instrumen-tation Setpoints Different setpoint methodology and different functional units 3/4. 1. 1 Reactivit Control S stem 3/4. l. 1. 1 Shutdown Margin-Tavg greater than 2000F Required shutdown margin is a function of boron concentration 3/4'. 1. 2 3/4. 1. 2. 2 Boration S stems Flow Paths-Operating Unique flow path configu-ration 3/4. 1. 2. 3 Charging Pumps-Operating Limited operation with common power supply

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3/4. l. 3 Reactivit Control S stem 3/4. 1. 3. 2

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Position Indication System-Operating STS modified for analog system instead of digital system 3/4.1.3.3 Position Indication System-Shutdown STS modified for analog system instead of digital system 3/4.2.1 Axial Flux Difference One target band instead of two and a changed monitoring time 3/4. 2. 2 Heat Flux Hot Channel Factor Additional limits for baseload and radial burn-down operation with augmented surveillances 3/4. 3. 1 Reactor Trip System Instru-mentation No response time instru-mentation, and different functional units and number of channels 3/4.3.2 Engineered Safety Feature Actuation System Instru-mentation No response time instru-mentation, different setpoint methodology, and different functional units and number of channels 3/4.3. 3 3/4. 3. 3. 1 Monitorin Instrumentation Radiation Monitoring For Plant Operations Different instrumentation 3/4. 3. 3. 2 Movable Incore Detectors Different utilization of detectors 3/4. 3. 3.3'ccident Monitoring Instrumenta-tionn Format change for Actions that are based on CTS 3/4.3.3.4 Fire Detection Instrumentation Fire watch patrol instead of instrumentation in turbine area 3/4.3.3.5 3/4. 3. 3. 6 Radioactive Liquid Effluent Monitoring Instrumentation Radioactive Gaseous Effluent Monitoring Instrumentation Different instrumentation Covers equipment in the design and adjusts actions to instrumentation in the design Relief Valves Valves not used for de-pressurization

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3/4. 4. 6 Reactor Coolant S stem Leaka e

Leakage Detection System Two detection systems instead of three 3/4.4. 6. 2 Operational Leakage Centrifugal charging pumps not used for safety injection 3/4. 4. 9 3/4. 4. 9. 3 Pressure/Tem erature Limits Overpressure Mitigating Systems Sharing by two units requires SI flow path isolation 3/4. 5. 1 Accumulators Instr umentation difference and CTS action time used 3/4. 5. 2 ECCS Subsystems Tavg R 350'F Subsystems not independent-centrifugal charging pumps not used for safety injection 3/4. 5. 3 ECCS Subsystems Tavg < 350 F

Safety injection is by RHR system only 3/4. 5.4 Refueling Mater Storage Tanks Two tanks with various flow path alignments t

3/4. 6. 1 3/4. 6. l. 6 Primar Containment Containment Structural Integrity SR modified for tendons that are not accessible at one end 3/4. 6. l. 7 Containment Ventilation System Design does not include mini-purge valves 3/4. 6. 2 3/4. 6. 2. 2 De ressur ization and Coolin

~Ss tems Emergency Containment Cooling System Three 50X capacity cooling units 3/4. 6. 3 Emergency Containment Filtering System Three 50X filtering units; non-independent 3/4. 6. 5 Combustible Gas Control-Hydrogen Monitors Flow path design requires additional surveillance 3/4. 6. 6 Post Accident Containment Vent System Design not consistent with STS SR 3/4. 7. 1 3/4. 7. 1. 3 Turbine C cle Condensate Storage Tank Main Steam Isolation Valves Two tanks shared by two units Action time is 50K of CTS time

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7. 2 Component Cooling Mater System Three 100K capacity pumps; non-independent loops 3/4.7.3

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Intake Cooling Mater System Three 100K capacity pumps; non-independent loops 3/4. 7. 4 Ultimate Heat Sink Minimum water level is not necessary 3/4.7.5 Control Room Emergency Ventilation System One system instead of two 3/4. 7. 8 3/4. 7. 8. 1 Fire Su ression S stems Fire Water Supply and Distribu-tion System Two water supplies instead of one 3/4. 7. 8. 2 3/4. 9. 7 Spray and/or Sprinkler Systems Crane Travel-Spent Fuel Storage Areas System shared by two units Crane interlocks and physical stops,not provided 3/4. 9. 8 4.9 ~ 8. 1 Residual Heat Removal and Coolant Circulation High Mater Level Water Level-Storage Pool SR for flow indicator was added Exceptions to LCO required for reracking and maintenance 3/4. 9 ~ 12 Handling of Spent Fuel Cask Consolidation of CTS requirements into STS format 3/4. 9. 13 Radiation Monitoring Consolidation of CTS requirements into STS format 3/4. 9. 14 Spent Fuel Storage Consolidation of CTS requirements into STS format

2. 4 RTS Based on 0 tions to the STS Ongoing programs to improve TS at nuclear power plants are continuing and result in options that can be used in lieu of STS until the options can be incorporated in a revision to the STS.

The following options have been incor-porated in the RTS:

RTS Basis 3.0.4, 4.0.3, 4.0.4 and Bases 3/4. 0 Generic Letter No. 87-09, "Section 3.0 and 4. 0 of the Standard Technical Specifications (STS) on the Applica-bility of Limiting Conditions for Operation and Surveillance Require-ments"

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4. 0.2 Generic Letter No. 89-14, "Line-Item Improvements in Technical Specifica" tions-Removal of 3.25 Limit on Extending Surveillance Intervals" Generic Letter 88-06, "Removal of Organizational Charts from Technical Specification Administrative Control Requirements" 2.5 Electrical TS g

r han the CTS.

Several of the interim TS have been renumbered or reoriented.

These changes are administrative and do not have safety implications.

The remaining interim TS are classified into two subgroups:

During the development of the RTS, the licensee requested that the conversion of the Electrical Power System CTS to the STS be deferred because of imminent plant modifications to improve the plant electrical

systems, and revised electri-cal TS would be needed at that time.

Rather than comment on the Proof and Review version of the

RTS, FPL proposed alternate Electrical Power Systems TS that differed from both the CTS and the STS.

As modified by staff review, these alternate TS were included in the Final Draft issued on March 14, 1989 and in FPL's amendment request of June 5, 1989.

This evaluation is based on the understanding that the electrical systems TS herein will be an interim TS.

The complete electrical systems TS will be re-evaluated by the staff against the STS during the review of the license amendment requesting approval of the electrical systems enhancements.

Therefore, this evaluation assures that the interim electrical TS maintain a degree of safety which is equivalent or reate t 1.

Changes that are equally or more conservative than the CTS in that they require equal or greater safety significant actions to be taken, or equal or faster action to be taken.

2.

Changes that allow more time between surveillances or before shutdown is required than allowed by the CTS, but which result in safer operation.

The following interim TS (subgroup

1) are equally or more conservative than the CTS and will improve plant safety.

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TS 3.8. 1.1 Actions a and b pertaining to inoperability of one star'tup transformer or one emergency diesel generator (EDG), respectively.

2.

TS 3.8.1.1 Action c pertaining to inoperability of one EDG during refueling surveillance.

3. =TS 4.8. 1. l. 1 pertaining to surveillance of the startup transformers.

4.

TS 4. 8. l.l. 2. a. 1),2) pertaining to verifying the minimum fuel volumes.

5.

TS 4.8. l. 1.2.d. 1) pertaining to verification of EDG ability to reject load.

6.

TS 4. 8. 1. l. 2. d. pertaining to inoperability of EDG trips during the emergency mode of operation.

7.

TS 4.8. l. 1.2.d.5)a pertaining to EDG loading surveillance.

8.

TS 4. 8. l. 1. 3 pertaining to reporting of EDG failures.

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9.. TS 3.8.1.2 and TS 4.8.1.2 pertaining to AC electric power limiting condition for operation (LCOs} and surveillance, respectively, for HODES 5 4 6.

10.

TS 3.8.2. lb pertaining to the LCO for inoperability of battery chargers.

ll. TS 3.8.2. 1 Actions a and b pertaining to inoperable battery chargers.

12.

TS 3.8.2.2 and TS 4.8.2.2 pertaining to the LCO, Action and surveillance for batteries and chargers for HODES 5 and 6.

13.

TS 3.8.3.1-pertaining to the LCD and Actions for inoperability of the 4160V and 480V AC buses.

14.

TS 3.8.3.2 and TS 4.8.3.2 pertaining to the LCO, Action and surveillance requirements for the 4160V and 480V AC buses for MODES 5 and 6'.

A number of TS changes (subgroup

2) are more conservative than the CTS in some

respects, but less conservative in other respects; the net effect is an overall enhancement in operational safety.

The CTS require that a controlled shutdown begin within one hour when an emergency diesel generator (EDG) and a startup transformer are both inoperable, or when two startup transformers are inoperable.

When one offsite source and one EDG are inoperable (TS 3.8. 1.1 Action d),

a controlled shutdown after 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sometimes required at recently built nuclear plants.

Because of the plant"specific startup transformer configuration at Turkey Point and because there are only two EDGs for both units at Turkey Point, the staff concluded that more time should be allowed to restore the startup transformer or EDG before requiring a shutdown.

The shutdown causes loss of power to the reactor coolant pumps and thus requires a cooldown on natural circulation and a challenge to some of the engineered safety features.

To ensure that the plant is in a safe condition during this period, the licensee agreed to begin reducing power on the affected unit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and to obtain a power level of 30K or less within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months if the affected unit is in MODE 1. If the unit is in MODES 2, 3 or 4, a controlled shutdown would begin after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

In addition, the interim TS would require other actions to demonstrate the operability of the other EDG, notify the NRC, verify the operability of the other startup transformer, demonstrate the operability of two out of five black start non-safety grade diesel generators and verify that they could be aligned to a 4160V safety bus in case of an emergency condition requiring such action.

If the startup transformer or EDG is not restored within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , both units would be required to begin a controlled shutdown.

However, if the EDG inoper-ability was due to a specific prolonged inspection activity during a shutdown, then 7 days would be allowed for beginning a controlled shutdown of the operating unit.

When two startup transformers are inoperable (TS 3.8.1.1 Action f), the STS require a controlled shutdown after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , as is recommended by Regulatory Guide 1.93.

The interim TS would similarly require a controlled shutdown after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ;

however, shutdown of the second unit is not required until the first unit is in cold shutdown.

The staff concludes it would be safer to complete the shutdown of the first unit before shutdown of the second because of the shared systems and equipment and the common control room.

When two EDGs are inoperable (TS 3.8. 1. 1 Action e), the STS require a controlled shutdown after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months as is recommended by Regulatory Guide 1.93.

The interim TS would also require a controlled shutdown after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

The interim TS allows 12-hours for both units to obtain hot standby, whereas the STS allows 6

hours.

A 12-hour period is appropriate at Turkey Point to allow a sequential and orderly shutdown of the two units which would produce less impact on the

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13 e

grid stability necessary to assure availability of offsite power sources to the facility.

The extension to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , has also been applied to other interim TS in the electrical system TS where action requires the shutdown of both units.

We find this to be acceptable.

The interim TS relax some of the EDG surveillance requirements (TS 4.8. l.1.2),

but also add r'equirements.

The additional requirements are consistent with the STS and therefore are acceptable.

The interim TS require fewer rapid start and rapid loading tests of the EDGs.

The test schedule frequency will be based on the last 20 valid tests rather than the last 100.

These relaxations are consistent with Generic Letter 84-15 and therefore are acceptable.

The load rejection test of the EDGs allows two seconds for the voltage to return to a specified maximum.

This is acceptable because the momentary transient will not damage any connected equipment.

The CTS do not specifically require both a startup and one-hour run of the EDG to demonstrate the operability of an EDG in accordance with the action statements.

However, the requirement for both is implied.

The run requirement is deleted from the interim TS.

This change is acceptable based on operating experience that has demonstrated that the EDGs need to be kept independent of disturbances on the non-vital and offsite power systems.

The CTS do not have action statements when certain 4160V or 480V emergency buses are inoperable.

When these buses are inoperable a controlled shutdown beginning within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is required in accordance with CTS 3.0. 1.

The interim TS (TS 3.8.3. 1 Action a) would allow 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months before a controlled shutdown.

This requirement is consistent with the STS and we therefore find it to be acceptable.

When one unit is shut down in a periodic refueling and maintenance outage and Motor Control Center (MCC)

D is inoperable, the interim TS (TS 3.8.3.1 Action d.) allows 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to begin a controlled shutdown.

The CTS do not have an action statement for this condition; therefore, TS 3.0.1 applies and requires a controlled shutdown beginning within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

The licensee states that the 24-hour period is consistent with the LCQs of the equipment powered by MCC D, and that the 24-hour period provides additional time to restore MCC D to operability without requiring a shutdown of the other unit (MCC D is common to both units).

We find the 24-hour period acce'ptable for the reasons stated.

3. 0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION NSHC DETERMINATION The licensee's request for amendments to the operating licenses for the Turkey Point Plant, Unit Nos.

3 and S, was noticed in the Federal

~Re ister on Oecember 5,

1989.

The staff received a request for hearing on these amendments.

That request is currently under consideration by an Atomic Safety and Licensing Appeal Board.

On May 15, 1990, the staff published in the Federal

~Re ister a

proposed determination of no significant hazards consideration (55 FR 20218).

No comments were received on that notice.

This is the staff's final determi-nation of no significant hazards consideration; there have been no significant changes to the proposed determination published earlier.

I The, amendments replace the current (custom) Technical Specifications (CTS),

which are part of the license issued in the early 1970 s, with a set of revised Technical Specifications (RTS) based on the staff's Standard Technical Specifications (STS) for Westinghouse-designed reactors.

The CTS and the RTS consist of 6 parts as follows:

0 S (l l

.r g 4

h I I

14 Part 1 - Definitions Part 2 - Safety Limits and Limiting Safety Settings Part 3 - Limiting Conditions for Operation (LCOs)

Part 4 '- Surveillance Requirements Part 5 Design Features Part 6 - Administrative Controls It should be nOted in reading the RTS that Parts.3 and 4 are presented as an integrated unit, so that the LCO and the surveillance requirement for a given plant system (or TS Section) are presented

together, system by system (or Section by Section).

The licensee's amendment application (the Application) submitted on June 5,

1989, as supplemented on July 12, 1989, November 3, 1989, February 13, 1990, May 1, 1990, June 21, 1990 'and July 20, 1990, included four attachments.

Attachment I includes the proposed RTS and revised Bases to support the RTS.

Attachment II is the licensee's safety evaluation, and Appendix A of Attachment II is a supplement to the safety evaluation and provides the No Significant Hazards Evaluation.

Attachment III identifies FSAR changes planned to keep the FSAR and RTS consistent with each other.

Attachment IV identifies certain safety improvements, in response to separate NRC initiatives, which will be implemented as a result of implementing the RTS.

Attachment IY was provided to assist the NRC in tracking progress on these other initiatives.

Throughout Appendix A of Attachment II to the Application, the licensee has characterized the proposed TS changes as:

(1) administrative (non-technical),

(2) more restrictive or more complete, (3) relaxations, and (4) deletion of selected requirements.

Administrative changes are non-technical in nature and are intended to make the TS easier to use for plant operations personnel.

More restrictive or more complete requirements are either more conservative than corresponding requirements in the CTS, or are additional restrictions which are not in the CTS.

The more restrictive or more complete requirements provide a safety enhancement.

Any relaxations of selected existing requirements are based on many reactor-years of operating experience in the nuclear reactor industry.

Requirements which are known to provide little or no safety benefit are counterproductive and may justifiably be eased or removed from the CTS.

In many cases the relaxed requirements already exist in the STS and have previously been issued in TS for other plants.

Deletion of selected requirements is described on page G-1 of Appendix A of Attachment II of the Application.

Oeletions consist of:

(1) requirements determined not to be needed for safety purposes, and (2) requirements which already exist in some other controlled document.

In the supplemental document submitted by the licensee on May 1, 1990, the changes evaluated for No Significant Hazards Consideration (Appendix A of Attachment II of the Application) are summarized and organized into tabular form.

The table, entitled "Categorization of Changes to the Current Tech Specs" is provided here for clarification as Table I.

The table includes no

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TABL CATEGORIZATIONOF C OTHE CURRENTTECH SPECS 0

CATEGORY INFORMATIONFROM NSH FOA REVISEO TECHNICALSPECIFICAllONS REVISEDTECH 1.0 1.17 2.1.1 2.1.2 2.2.1 3/4.0 0.1 and Mode Iss 3/4.1.1.1 3/4.1.1.2 314.1.1.3 3/4.1.1.4 314.1.2.1 3/4.1.2.2 3/4.1.2.3 3/4.1.2 4 3/4.1.2.5 3/4.1.2.6 3/4.1.3.1 3/4.1.3.2 3/4.1.3.3 3/4.1.3.4 3/4.1.3.5 3/4.1.3.6 3/4.2.1 3/4.2.2 3/4.2.3 3/4.2.4 3/4.2.5 3/4.3.1 3/4.3.2 CURRENI TECH.

SPEC.

1.0 ANDTABLE 4.1-1 1.4 3.05 and 83.05 1.1 2.1 and 82.1 1.1 2.2 and 82.2 2.3 3.0 4.0 and 83.0 Thto hout S 3.2.1.1, 3.2.4.c, Table 4.1-2 Item 1.e 4.11 and 6.9.3.m Table 4.1-2 ilem 1.e 3.1.2.1 None Adds New S ci5cation 3.6.a, Table 4.1-1 item 19, and Table 4.18-1 item 8 3.6.a, 3.6.b.2, 3.6.b.4, 3.6.c.2, 3.6.c 4, 3.6.d.2, and Table 4.18-1 item 8 3.6.b.1, 3.8.c.1, 3.6.d.1, and Table 4.1-1 items 12 and 16 Table 4.1-1 ilom 14 and Table 4.1-2 item 3 3.4.1.a.1, 3.6.b.3, 3.6.b.6, 3.6.c.3, 3.6.c.6. and Table 4.1.2 items 2 and 3 3.6.b.5, 3.6.c.5, 3.6.d, and Table 4.18-1 item 8 3.2.2, 3.2.4a, 3.2.4b, 3.2.5 and Table 4.1-2 item 5 3.2.5. and Tabb 4.1-1 items 9 and 10 Table 4.1-1 item 9 3.2.3 and Table 4.1-2 item 5 3.2.1.a 3.2.1.b 3.2.1.c 3.2.1.d 3.2.1 3.2.6.c Ihru 3.2.6 and 3.2.8 3.2.6.a, 3.2.6.b, and Table 4.1-1 item tb 3.2.6.a, 3.2.6.b, and TaMe 4.1-1 item 1b 3.2.6h and 3.2.6i 3.1.6 3.5.1 and Table 4.1-1 3.5, Table 3.52, Table 3.5-3, Table 3.5.4 and Table 4.1-1 NSH APPENDIXAPAGE REFERENCE 1-1 THRU 1.3 1-4 THRU 1.7 2-1 THRU 2.2 2-3 THRU 24 2-5 THAU 2-7 314 0.1 THAU 3/4 0-4 3/4 0-5 THRU 3/4 0-7 3/4 1-1 THRU 3/4 1A 314 1-5 THRU 3/4 1.6 3/4 1.7 THAU 3/4 1-8 3/4 1.9 THAU 3/4 1-10 3/4 1-11 THRU 3/4 1-13 3/4 1-14 THRU 3/4 1-1$

3/4 1-19 THAU 314 1-2I 3/4 '1-22 THAU 3/4 1-24 3/4 1-25 THRU 3/4 1-29 3/4 1.30 THRU 3/4 1-32 3/4 1.33 THRU 3/4 1-36 3/4 1-37 THRU 3/4 1-39 3/4 1.40 THRU 3/4 1-41 3/4 1-42 THRU 3/4 1-43 314 1-44 THRU 3/4 1-45 314 1-46 THAU 314 1-47 3/4 2-1 THRU 3/4 2-3 3/4 2.4 THRU 3/4 2-5 3/4 2-6 THRU 3/4 2-8 3/4 2.9 THRU 3/4 2-12 3/4 2.13 THAU 3/4 2-14 3/4 3-1 THAU 3/4 3.5 314 3 6 THRU 3/4 3.9 ADMINISTRATIVE CHANGES A.2)a.1,2,3,4,5,6, A.2 a.7 89 1011 A.2 A.2.a A.2.a A.2.a A.2.a A.2.a A.2}.a A.2.a A.2.a A.2.a A.2).a A.2).a A.2).a A.2).a A.2).a A.2).a A.2).a A.2).a A.2.a A.2.a A.2.a A.2.a A.2.a A.2).a A.2).a A.2.a A.2.a A.2.a A.2).a MOREIXMPIHEOR RESTRICTIVE CHANGES A.2).b.

A.2 A.2 b.1 2 A.2 b A.2 b.1 2 A.2 b.1 A.2)b.1,2,3 A.2 b.1 2 3 A.2 b A.2 b A.2)b.1,2 A.2)b.1.2 A.2)b.1,2,3 A.2)b.1,2,3 A.2)b.1,2,3 A.2}b.1,2 A.2)b.1,2,3 A.2)b.1,2 A.2 b.1 2 3 A.2 b.1 2 A.2 b.1 2 3 A.2 b.1 2 A.2 b.1 2 A.2)b.

A.2 b.l 2 3 4 A.2 b.1 2 3 A.2 b.1 2 A.2)b.1,2,3,4 CHANGES THATAAE REIAXATIONS A.2 A.2.c A.2.c.1 2 A.2 b A.2).c.1,2,3 A.2).c A.2).c.1,2,3,4,5 A.2).c.1,2,3 A.2).c.1,2 A.2).c.1,2,3 A.2).c.

A.2).c.1,2,3,4,5 A.2).c.1,2 A.2.c A.2.c A.2)b.

A.2).c.

A.2.c.l 2 A 2.c I 2 3 4 5 A.2).c.1,2,3 OELETIONS FROMTHE CURRENT TECH SPECS A.2)c.1,2,3,4,5,6,7,8 A2c91011 1213 Pago 1

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CATEGORIZATIONOF C THE CURRENT TECH SPECS CATEGORY INFOANATIOIIFROM NSH FOR REVISED TECHNICALSPECIFICATIONS 0 CURRENI'TECIL SPEC.

NSH APPENDIX APAGE AEFE ADMINISTRATIVE MOf%COMf IEIEOR CIQNGES RESTRICTIVE CHANGES CIIANGESTHATARE RELAXATIONS DELERCNS FROM THE CUAAENTTECHSPECS 3/4.3.3.1 314.3.3.2 3/4.3.3.3 314.3.3.4 3/4.3.3.5 3/4.3.3.6 3/4A.1.1 314.4.1.2 3/4A.1.3 3/4A.1 A.I, 3/4A.IA.2 3/4.4.2.1 3/4.4.2.2 3/4.4.3 314.4.4 314.4.5 314.4.6.1 3/4A.6.2 3/4.4.7 3/4.4.8 3/4.4.9.1 314 A.9.2 3/4.4.9.3 3/4.4.10 3/4.4.11 3/4.5.1 Table 3.5.3 item 4, Table 3.5.4 item 10. Table 4.1-1 items

18A, 18B, 38a. 38b, and Table 3.5.5 items 13a and 13b 3.2.7 Table 3.5.5 items 1 thru 11 and 13 thtu 15, Tablo 4.1-1 items 6, 15A, 158, 16, 17A, 178, 26, 27, 28, 29, 30, 34, 35, 36, 37, 38 39 and 40 3.14.1 and 4.15.1 3.9.1.C. Table 3.9-2, and Table 4.1-3 3.9.2.C, TaMe 3.9-3, Table 3.9.4 and Table 4.1-4 3.1.1.a.l, 3.1.1a.3. 3.1.1.a.4, 3.4.lc and Table 4.1-2 item 18 3.l.l.a.2, 3A.l.d, and Table 4.1-2 1.23, 3.1.1.a.2, 3.1.$.a.5, 3.4.l.e and Table 4.$ -2 item 18 3.1.1.a.2, 3.1.1.a.5. 3.4.l.e, and Table 4.$.2 item 18 3.1.l.a.2. 3.4.1.e, and Table 4.1-2 item

$ 8 3.1.l.c.l, TaMe 4.$ -2 item 6, and 83.$.1

+$.$.c.2, Tablo 4.1-2 ilom 6, and 83.1.1 3.$.1.d 3.1.1.e.l 3.1.1.e.2 and 3.1.1.e.3 4.2.5 3.1.3.1 B3.$.3 Table 4.1-1 item 20 3.1.3a, 3.1.3b, 3.1.3c, 3.1.3d, 3.$.3e. 3.1.3g, 3.16, 4.17, and Table 4.1.2 item 11 3.1.5 and TaMe 4.1-2 itom lb 3.1.4, 83.1.4, and Table 4.1-2 item I 3.1.2 83.1.2 4.20 and 84.20 3.1.2 and 83.$.2 3.15 4.$ 6 83.$ 5 and 84.$ 5 4.2 and 4.3 3.1.1.$

4.$ 9 83.1.1 and 84.19 3A.l.a.3, 3A.l.b.l, 4.5.2.b.3, Table 4.1-1 item 21, and Table 4.1-2 item 10 3/4 3.10 THRU 3/4 3-13 3/4 3-$ 4 THflU 3/4 3-15 3/4 3-16 THRU 314 3-19 314 3-20 THRU 3/4 3-22 3/4 3-23 3/4 3-24 THAU 3/4 3-26 3/4 4-1 THAU 3/4 4.3 314 4-4 THAU 3/4 4.5 314 4-6 THRU 314 4.7 3/4 4 8 THRU 3/4 4.10 3/4 4-11 THRU 3/4 4-13 314 4-14 THRU 314 4-17 3/4 4-18 THAU 3/4 4.19 3/4 4.20 THAU 3/4 4.21 3/4 4 22 THAU 314 4.24 3/4 4.25 THRU 3/4 4 26 314 4.27 THAU 314 4.30 314 4-31 THAU 3/4 4-34 3/4 4.35 THAU 3/4 4-37 314 4.38 THRU 3/4 4 40 314 4.41 THAU 314 4 42 314 4.43 THAU 3/4 4 44 314 4-45 THAU 3/4 4 4?

3/4 4 48 THRU 3/4 4.49 3/4 4-50 THAU 3/4 4 51 3/4 5.1 THRU 3/4 5-3 A.2).a A.2.a A.2).a.'I, 2 A.2.a A.2).a A.2).a A.2).a A.2).a A.2).a h.2).a A.2).a A.2).a A.2).a A 2.a A.2.a A.2.a A.2.a A.2).a A.2.a A.2).a A.2.a A.2.a h.2.a A.2.a A.2.a A.2).a A.2)b.1,2,3,4,5,6 A.2 b.l 234 A.2)b.1,2 A.2 b.l A.2)b.1,2,3,4,5.6,7 A.2)b.

A.2)b.

A.2)b.1,2,3,4 A.2)b.'1,2,3 A.2 b.l 2 3 A.2 b A.2 b.l 23 A 2 b I 23 A.2)b.1,2,3,4,5 A.2 b.l 2 A.2)b.1,2 A.2 b.$ 2 A.2 b.l 2 A.2 b.$ 2 3 4 A 2 b A.2)b. $,2,3,4,5 A.2).c.1,2 A.2).c.1,2 A.2.c.l 2 h.2).c.

A.2).c.1,2,3 A.2).c.1,2 A.2).c.1,2,3 A.3.a. b A.2.c A.2).c.1,2,3 h.2.c.

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CATEGORIZATIONOF C TO THECURRENT TECH SPECS CATEGORY INFORMATIONFROM NSH FOR REVISED TECHNICALSPECIFICATIONS REVISEDTECH.

SPEC NO.

CURAENFTECH.

SPEC.

S NSH APPENDIXAPAGE REFERENCE S ADMINISTRATIVE MOf%004PU'TEOR CHANGES THATARE CHANGES RESTRICllVE CHANGES RE IAXATKNS DEM'KNSFROMTHE CURRENT TECHSPECS 3/4.5.2 3/4.5.3 3/4.5.4 314 6.1.1 3/4.6.1.2 3/4.6.1.3 314.6.1.4 3/4.6.1.5 3/4.6.1.6 3/4.6.1.7 3/4.6.2.1 3/4.6.2.2 3/4.6.3 3/4.6 4 3/4.6.5 3/4.6.6 314.7.1.1 3/4.7.1.2 314.7.1.3 314.7.1.4 3/4.7.1.5 314.7.'I.8 314.7.2 3/4.7.3 3/4.7.4 3/4.7.5 3/4 7.6 3/4.7.7 3/4.7.8.1 314.7.8.2 3/4.7.8.3 314.7.8.4 314.7.9 3/4.8.1.1 3.4.1.a 4 thru 3 4.1.a.?,

3.4,1.b.2, 3.4.l.b.4 thru 3 4.1.b.7, 4.5.1, 4.5.2.a.

4.5.2.b.l, 4.5.2.b.2, 4.5.2b.4, and Table 4.18 1 items 1 and 2 None Adds Now S ci6cation 3.4.1.a.l and Table 4.1-2 item 2 3.3.1 4.4.1 4.4.2 and 44.3 3.3.4 and 4.4.2 3.3.2 None Adds New S cilication 4.4.5 4.4.6 and 4.4.7 3.3.3 and 4.4.2 3.4.2 4.6 Table 4.18-1 item 4 3.4.2 and 4.6 3.4.3a 3.4.3b and 4.7 3.3.3, 83.3.3, 4.4.3, and Table 4.2-1 item 8 84.18, Table 3.5-5 item 12, Table 4.1-1 item 37, and Table 4.18-1 item 11 3.4.8, 4.7.2, and Table 4.18-1 item 9 3.8.1a 83.8 Table 4.1-2 itom 7 3.18 4.10 Table 4.18-1 item 3 3.19 83.19 and 4.22 3.8.2 3.8.1.b, 3.8.1.c, 3.8.3, 4.9, and 84.9 3.20 4.21 83.20 and 84.21 3 4.4, 83.4.4, and Table 4.18-1 item 6 3.4.5, 83.4.5, and Table 4.18-1 item 7 None Adds New S cilication 3.4.7 4.7.3 and 84.7 3.13 and 4.14 3.11 4.13 and 83.11 3.14.2 and 4.15.2 3.14.3 and 4.15.3 3.14.4 and 4.15.4 None Adds New S cilication 3.14.5 and 4.15.5 3.7, 4.8.1, 83.7. 84.8.

Table 4.8-1, Table 4.1-2 item 12 and Table 4.18-1 item 5 3/4 5-4 THRU 3/4 5.9 3/4 5-10 THRU 3/4 5.11 3/4 5-12 THRU 3/4 5-13 3/4 6-1 THAU 3/4 62 3/4 6-3 THRU 3/4 6 4 3/4 6.5 THRU 3/4 6 6 3/4 6-7 THRU 3/4 6 8 314 6-9 THRU 314 6.10 3/4 6-11 THAU 3/4 6.12 3/4 6-13 THAU 3/4 6-14 3/4 6-15 THAU 3/4 6-18 3/4 6-19 THAU 3/4 6-21 3/4 6-22 THRU 314 6-24 3/4 6-25 THAU 3/4 6-26 3/4 6-27 THRU 3/4 6-29 3/4 6-30 THRU 3/4 6-31 3/4 7-1 THRU 3/4 7-3 3/4 7-4 THRU 3/4 7 8 3/4 7-7 THAU 3/4 7.8 3/4 7-9 THAU 314 7-10 3/4 7-11 THAU 3/4 7-12 3/4 7-13 THAU 3/4 7-14 3/4 7-15 THRU 3/4 7-18 3/4 7-19 THAU 3/4 7-22 3/4 7-23 THRU 3/4 7-24 314 7-25 THRU 3/4 7-27 314 7-28 THAU 3/4 7.29 314 7-30 THRU 3/4 7-32 3/4 7-33 THAU 3/4 7-35 314 7-36 THRU 3/4 7-39 3/4 7.40 THAU 314 ?-42 3/4 7-43 THRU 314 7-44 3/4 7-45 TH U 3/4 7-47 3/4 8-1 THRU 3/4 8-9 A.2).a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2).a A.2).a A.2).a A.2.a A.2.a A.2.a A.2.a A.2).a A.2.a A.2).a A.2).a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2).a.l A.2)b.1,2,3,4,5,6 A.2 b.l 2 3 A.2 b.l 2 3 A.2 b.l 2 3 A.2 b.l 2 3 A.2 b.l 23 A.2 b.

A2b.12345678 A.2 b.l 2 3 4 5 A.2 b.l 2 3 A.2 b.l 2 A.2 b.l 2 3 4 A.2)b.1 A.2)b.1,2,3 A.2)b.

A.2 b.l 2 A.2 b.

A.2 b.l 23 A.2)b.1,2 A.2 b A.2)b.1,2,3 A.2)b.1,2 A.2 b.l 23 4 A.2 b.

A.2 b.l 2 A.2 b.l 23 A.2 b.l 2 3 A.2 b.l A.2)b.1,2,3,4.5.6,7,8 A.2)b.9,10,11,12 A.2).c.1,2,3,4,5,6 A.2.c A.2.c None A.2.c. 1 2 3 A2.c 1 23 A.2.c A.2).c.

A.2).c.

A.2.c.l 2 A.2.c A.2 b None A.2).c.

A.2).c.1,2,3,4 A.2).c.1,2,3,4 A.2.c.l None A.2.c A.2.c.l 2 A.2.c,l 2 A.2 b.l 2 A.2.c.l 2 A.2).c.1,2,3,4,5,6,7 Pago 3

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CATEGORIZATIONOF C ESTO THE CURRENTTECH SPECS CATEGORY INFORMATIONFROM NSH FOR REVISEO TECHNICALSPECIFICATIONS 3/4.8.1.2 314.8 2.1 3/4.8.2.2 3/4.8.3.1 3/4.8.3.2 314.9.1 3/4.9.2 3/4.9.3 3/4.9.4 3/4.9.5 3/4.9.6 3/4.9.7 314.9.8.1 3/4.9.8.2 3/4.9.9 314.9.10 3/4.9.1 1 3/4.9.12 314.9.13 3/4.9.14 314.10.1 314.10.2 3/4.10.3 314.10.5 3/4.11.1.1 3/4.11.1.2 314.11.1.3 3/4.1 1.2. 1 3/4.11.2.2 3/4.11.2.3 3/4.11.2.4 3/4.1 1.2.5 3/4. 1 1.2.6 3/4.1 1.3 314.1 1.4 3/4.'I 2.1 314.12.2 314. 1 2.3 5.1 CUAAENfTECH.

SPEC.

S None Adds New S ecitication 3.7 4.8.2 83.7 and 84.8 None Adds Now S cilication 3.7 83.7 Table 4.16-1 item 10 None Adds New S cilication 3.10.8, 83.10.8, and Table 4.1-2 item 13 3.10.3, 83.10.3, and Table 4.1-1 item 3 3.10.5 and 83.10.5 3.10.1 and 83.10.1 3.10.8 and 83.10.6 Table 4.1-2 item 9 3.10.9 and 83.10.9 3.10.7.1, 83.10.7, Table 4.1-1 item 13, and Tabte 4.1-2 item 18 3.10.7.2, 83.10.7, and Table 4.1-2 item 18 3.10.2, 83.10.2, and Table 4.1-2 item 8 None Adds New S cification None Adds New S cilication 3.12 83.12 Table 4.1-2 item 17 3.10.4, 83.10.4, and Table 4.1-1 item 18A 3.17 83.17 Table 4.1.2 item 13 3.2.1l 3.2.1a, 3.2.1b, 3.2.1c, 3.2,6.d, and 3.2.6.h 3.1.2.1 3.2.1.a 3.2.1.b 3.2.1.c None Adds Now S cification 3.9.1a 83.9.1a and Table 3.9-1 3.9.1.b 6.9.3e and 83.9.1.b 3.9.1.d 83.9.1.d and 6.9.3.1 3.9.2a 83.9.2.a and Tabte 3.9.3 3.9.2b 83.9.2b and 6.9.3.e 3.9.2.c 83.9.2.c and 6.9.3.e 3.9.2e 83.9.2e and 6.9.

3.9.2.

83.9.2.

TaMe 3.9-4 3.9.2.t and 83.9.2.I 3.9.3 and 83.9.3 3.9.2.h 83.9.2.h and 6.9.3.h 4.12.1 84.12.1 and 6.9.3i 4.12.2 and 84.12.2 4.12.3 and 84.12.3 5.1 NSH APPENDIXAPAGE REFER&CE 3/4 8.10 3/4 8-11 THAU 3/4 8-14 3/4 8-15 3/4 8-16 THRU 3/4 8.18 3/4 8-19 314 9-1 THRU 3/4 9-2 3/4 9-3 THRU 3/4 9-4 3/4 9.5 THRU 3/4 9 6 3/4 9-7 THRU 3/4 9.8 3/4 9.9 THRU 3/4 9-10 3/4 9-11 THRU 3/4 9.12 3/4 9-13 THRU 3/4 9-14 3/4 9-15 THRU 3/4 9-'l7 3/4 9-18 THRU 3/4 9-20 3/4 9-21 THRU 3/4 9-22 3/4 9.23 THAU 314 9-24 3/4 9-25 THAU 3/4 9-26 3/4 9-27 THRU 3/4 9-28 3/4 9-29 THRU 314 9-30 3/4 9-31 THAU 3/4 9-32 3I4 10-1 THRU 314 10-2 3/4 10-3 THRU 314 10-4 3/4 10-5 THAU 3/4 10.6 314 10-7 THAU 314 10 8 3/4 11-1 THAU 3/4 11-2 3/4 11-3 THRU 3/4 11-4 3/4 11-5 THAU 3/4 11-6 314 11-7 THRU 314 11.8 3/4 11-9 THRU 3/4 11-10 3/4 11-11 THAU 3/4 11-12 3/4 11-13 THAU 314 11-14 3/4 11-15 THAU 3/4 11-16 3/4 11-17 THAU 3/4 11-18 3/4 11-19 THAU 3/4 11-20 3/4 11-21 THAU 314 11-22 3/4 12-1 314 12-2 314 12-3 THAU 314 12-4 5-1 THRU 5-3 ADMINISTRATIVE CHARLES A 2.a A.2.a A.2.a A 2.a A.2.a A.2).a A.2).a A.2.a A.2.a A.2.a A.2.a A.2.a A.2).a A.2).a A.2).a A.2.a A.2.a A.2.a A.2).a A.2.a A.2.a A.2).a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a MORECCMPlEIEOR RESTAICllVECHANGES A.2 b.1 2 3 A.2 b.1 23 4 56 A.2)b.1,2 A.2)b.1,2 A.2 b A.2 b A.2 b A.2 b.1 2 3 4 5 A.2 b A.2)b.

A.2)b.

A.2 b A.2 b A.2 b A.2)b.1,2 A.2 b A.2 b A.2)b.

A.2 b A.2 b.

A.2 b A.2 b A.2 b.

CHANGES THATARE AELAXATIONS A.2.c.1 2 A.2.c.1 A.2).c.

A.2.c A.2).c.

A.2).c.

None A.2 b None A.2.c.

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CATEGORIZATlONOF TOTHE CURRENT TECH SPECS CATEGORY INfORMATIONFROM NSH FOR REVISED TECHNICALSPECIFICATIONS CURRENI'TECH, SPEC.

NSH APPENDIXAPAGE REFERENCE ADMINISTRATIVE hhOfKCQMPLEIEOR CHANGES THATARE CHANGES RESTRICTlVE CHANGES RELAXATlONS DEIETIONSFROM THE CURRENr TECHSPECS-.

5.2.1 5.2.2 531 5.3 2 5.4.1 5.4 2 5.5 5.6.1 5.6 2 5.6.3 5.7 61 6.2.1 6.2.2 6.2.3 6.3 6.4 6.5.1 6.5 2 6.5.3 6.6 6.7 6.8 6.9.1 6.9.2 6.10 6.11 6.1 2 6.13 6.14 6.15 None Adds New Desi n Feature 5.3.A.2 5.2.1 5.2.5 None Adds New Desi n Feature 5.2.3 None Adds New Desi n Feature 5.4.2 and 54.3 None Adds New Desi n Feature None Adds New Desi n Feature None Adds New Desi Feature 6.2.1 6.2.2 6.3.1 6.3 64 6.5.1 6.5.2 None Adds New S cification 6.6 6.8 6.13 6.14 6.15 and 6.16 6.9 6.9.1 6.9.3 and 6.9.4 6.9.3.a 6.9.3.b and 6.9.3.c 6.10 6.1 1 6.12 6.1 7 6.18 None Adds New S cilication 3 4.1.a.2 Table 4.1-1 Table 4.1-2 4.18 Section 6 Design Features 5-4 5-5 THRU 5.6 5-7 THRU 5.8 5.9 THRU 5-10 511 THRU 5.12 5-13 THRU 5-14 5-15 THRU 5.16 5-17 THRU 5-18 5.19 THRU 5 20 5-21 THRU 5-22 5-23 THRU 5-24 6.1 THRU 6.2 6-3 THRU 64 6.5 THRU 6.6 6-7 THRU 6.8 6.9 THRU 6.10 6-11 THRU 6-12 6.13 THRU 6 15 6.16 THRU 6-17 6-18 THRU 6.19 6.20 THRU 6.21 6-22 THRU 6 25 6-26 THRU 6.28 6-29 THRU 6-32 6-33 THRU 6 34 6-35 THRU 6-36 6-37 THRU 6-38 6.39 THRU 6-41 6.42 THRU 6.43 6-44 THRU 6-45 6-46 THRU 6.47 G-1 THRU G.4 G-1 THRU G-4 G-1 THRU G-4 G-1 THRU G-4 G.1 THRU G-4 A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A,2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a h.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2.a A.2 b A.2 b.1 2 A.2 b A.2 b A.2 b.

A.2 b A.2 b.1 2 3 A.2.a A.2 b.2 A.2 b.1 2 A.2 b.1 2 A.2 b.1 2 A.2 b.1 2 A.2 b.

A.2.a Norre Norw h.2 c.1 A.2 b A.2.a A.2 b.l and A.2 c.1 2 A.2 c A.2 b.3 one item seven items four items one item eleven items 3.4.t.a.2 Table 4.1-1 items, 12, 13 14 16 19 24 25 Table 4.1 2 rtems 4, 9

15 and 16 4.18 wer availabilit 5.2 2a 8 2b. 5.3.A.1, 5.3.A.2, 5.3.81 8 82, 5.3.C.1, 5.3.C.2, 5.3.C.3, 5.4.1 and 5.2.4 Page 5

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new information, but lists and organizes all the changes.

The first column of the table lists the new RTS section that results from the changes to the CTS.

The second column of the table lists the CTS sections being changed.

The third column lists the page reference in Appendix A of Attachment II of the Application where the changes are described and evaluated for No Significant Hazards Considerations.

The last four columns list the changes, by category, using the notation from Appendix A.

The Commission provides bases for a no significant hazards consideration determination in 10 CFR 50.92.

These include the three standards set forth in 10 CFR 50. 92 for determining whether a significant hazards consideration exists.

Under the Commission's regulations in 10 CFR 50.92, a proposed amendment involves no significant hazards consideration if operation of the facility in accordance with the proposed amendments would not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

3. 1 NSHC Evaluation The licensee performed a detailed evaluation of the changes proposed in the RTS against the above standards and concluded that none of the proposed changes involves a significant hazards consideration.

Reference:

Attachment II, Appendix A, of the Application.

The staff reviewed the No Significant Hazards Evaluation (NSHE) provided in Attachment II, Appendix A to the June 5, 1989 license amendment proposal.

Based on that review, the staff agrees with the licensee's conclusions and has made a final determination that the proposed amendments involve no significant hazards considerations.

The staff has selected examples of the proposed TS changes in each of the four categories of characterization (administrative, more restrictive, etc.)

employed by the licensee and which also cover the six parts of the

RTS, and they are discussed below.

These examples are considered to be typical of the proposed changes.

The staff's evaluation of no significant hazards is presented below.

3. 2 Cate or 1 - Administrative chan es Examples of administrative changes include consolidation of requirements in one

place, reformatting of requirements, numbering of all pages, and revision of definitions (Part 1 of the TS).

Three examples are discussed below.

~Exam le 1:

Consolidation In the CTS, requirements for a given plant system or component are often dispersed throughout a number of sections of the CTS.

The RTS consolidates the requirements for a given system or component into one section, which improves TS organization.

The changes in the Refueling Water Storage Tank requirements are an example of this type of change.

The CTS limits on borated water volume and boron concentra-tion are located on page 3.4-1 of the CTS in Section 3.4. l.a.l, and the require-ment for a weekly verification of boron concentration is found in Table 4.1-2, item 2 (there is no page number; the table is located six pages past page 4.1-1).

In the RTS, this information is consolidated in one place in a Refueling Water Storage Tank TS, on page 3/4 5-9, located in Section 3.5.4, Limiting Conditions for Operation, and on the same

page, Section 4. 5.4, Surveillance Requirements.

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16 Because there is no technical change related to consolidation, i.e., the consolidated requirements remain the same, consolidation does not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The three standards of 10 CFR 50.92 are satisfied and the staff has made a final determination that there are no significant hazards considerations.-

The staff also made a final determination that, throughout the RTS where con-solidation has been made, there are no significant hazards considerations involved.

~Exam le 2:

Reformatting In the CTS, beginning on page 3.1-1, the format of the TS consists of a very generalized statement of applicability, a statement of purpose of the TS and a detailed specification which combines requirements with actions to be taken if requirements are not met.

The RTS have an improved format which sets forth the requirement stated as the Limiting Condition of Operation (LCO), the operational mode applicability, and the statement of action required if the LCO is not met.

These requirements (LCO, applicability, and action) are organized as separate entities and presented in the same sequence with a heading in capital 'letters throughout Sections 3 and 4 of the RTS.

Reformatting'as not resulted in any changes to the plant operating requirements that are in the RTS.

Since reformatting does not change any of the requirements contained in the TS, reformatting does not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The three standards of 10 CFR 50.92 are satisfied, and the staff has made a final determination that there are no significant hazards considerations.

The staff has also made a final determination that reformatting throughout the RTS involves no significant hazards considerations.

~Exam le 3:

Revision of Definitions In its no significant hazards evaluation (Appendix A of Attachment II of the Application, pages l-l through 1-3), the licensee evaluated changes in Part 1 (definitions) of the CTS and concluded that no significant hazards consideration is involved.

On page l-l of Appendix A, in items 2. a,

2. b and 2. c respectively, the licensee notes that 11 new definitions have been added, refueling interval has been explicitly defined, and 13 definitions have been deleted.

Generally, the added definitions are related to specific parameters which the TS help to

control, such as various leakages,

tests, and neutron flux.

Examples of added definitions include pressure boundary leakage, actuation logic test, and axial flux difference.

The deleted definitions, on the other hand, are general terms which are either not needed for specific controls, or are not needed because they already exist somewhere else (for example, in the Code of Federal Regulations).

Examples of deleted definitions include design power, safety limits, and reactor protection system.

The licensee has addressed the three criteria of 10 CFR 50.92(c) and determined that they are satisfied because:

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17 0

"(1) The proposed change as described in Item 2.a is similar to example (i) of 48 FR 14870 in that it is an administrative change which consolidates current requirements into a technical specification format consistent with the Standard Technical Specifications and does not involve ['technical3 or plant modifications.

(2)

The change in Item 2.b is similar to example (ii) of 48 FR 14870 in that -it provides additional restrictions and controls by requiring surveillances with frequency "R" to be performed at least once per 18 months.

(3)

The proposed changes described in Item 2. c represent definitions of terms which are not used or which are defined in other places in the revised technical specifications.

In some cases, the proposed changes described in Item 2.c represent restrictions to plant oper ation.

In each case where an omitted definition contains a restric-tion, the restriction is included in another section of the revised technical specifications.

Therefore, the proposed changes described in Item 2.c also are similar to example (i) of 48 FR 14870 in that they are administrative changes which consolidate current requirements into a technical specification format consistent with the Standard Technical Specifications and do not involve technical or plant modifications."

'he staff agrees with the licensee's conclusion that there are no significant hazards considerations, with the following additional comments.

The changes in definitions described on pages l-l through 1"3 of Appendix A to Attachment II t

of the Application do not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possi-bility of 'a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety because:

(a) the added definitions help to clarify and avoid misinterpretation of existing terms related to specific controls and tests, and (b) those definitions deleted are either very general and therefore not very useful, or they exist elsewhere, or their useful content is included in the specific relevant technical specifi-cation.

For example, one deleted definition is Limiting Condition for Operation (LCO).

While the LCO definition might be of interest to persons outside the reactor operations field, the definition in the CTS ("those restrictions on reactor operation, resulting from equipment performance capability, that must be enforced to ensure safe operati'on of the facility") is obvious and unnecessary for reactor operators and personnel and NRC staff by whom the TS are used.

Furthermore, LCO is described in 10'CFR 50.36.

An example where a deleted definition has its useful content transferred to a specific TS is "Safety Limits".

In.the CTS on page l-l, the definition states'that if any safety limit is

exceeded, the reactor shall be shut down until the AEC (now the NRC) authorizes resumption of operation.

This statement refers to action required rather than stating a definition.

The action to shut down has been transferred to the individual RTS section for safety limits in Part 2.

The action to remain shutdown until NRC approval is obtained to restart is transferred to page 6-13 of the RTS in Section 6.7. l.d.

The remainder of the "Safety Limits" definition in,.the CTS is general in nature and is also described in 10 CFR 50.36.

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18 The revision of the definition of refueling interval specifies a time interval of 18 months or less, which clarifies what is meant by refueling interval.

No specific time interval is defined in the CTS so that a refueling interval could be any convenient period of time resulting from an actual fuel cycle.

The revised definition brings the Turkey Point definition in line with STS practice in the industry.

No significant increase in probability or consequences of an

accident, creation of a new or different kind of accident, or reduction in a margin of safety can result from these changes because the revised'efinition is narrower, and is thus encompassed within the CTS definition.

Therefore, the staff has made a final determination that the revisions of the definitions involve no significant hazards considerations.

3.3 Cate or 2 " RTS Re uirements Which Are More Restrictive or More Com lete Than CTS Re uirements Examples of proposed changes in requirements which are more restrictive or more complete than those now in the CTS are discussed below.

These include:

examples of changes to safety limits and limiting safety settings in Part 2 of the TS, examples of changes to LCOs in Part 3 of the TS, examples of changes to sur-veillance requirements in Part 4 of the TS, and examples of changes to administrative controls in Part 6 of the TS.

~Exam le 1:

Safety limits and limiting safety settings In CTS Sections 1.1 (page l-l) and 2.1 (page 2.1"1) covering reactor core safety limits, the combination of reactor pressure, temperature, and thermal power level are not permitted to exceed certain limits provided in Figure 2. 1-1 (no page number).

However, no explicit required action is identified in the CTS if the limits are exceeded.

Instead, the operators are referred to CTS Section 3.0. 1 which requires that action be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to reduce reactor power and place the reactor in a different approp~iate operating mode,,

the earliest being Hot Standby (operational mode 3) within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

However, in the RTS Section

2. 1 (page 2-1), the same limits apply, but a specific explicit action statement has been added.

This follows the format of the STS.

The action statement is more restrictive because (1) it requires the operators to place the reactor in Hot Standby within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months instead of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and (2) it refers the operator to Section 6.7. 1 (page 6-12} of the RTS, which provides reporting requirements.

Also, CTS Section

2. 1 requires that specified power/pressure/temperature limits for one-and two-loop operation and natural circulation not be exceeded.

However, because these limits have not been analyzed in the safety analysis, they are being deleted.

In the licensee's no significant hazards evaluation, Appendix A of Attachment II of the Application, pages 2"1 and 2-2, the licensee evaluated the changes to Section

2. 1 of the CTS in accordance with the three standards of 10 CFR 50.92(c) and concluded that the changes do not involve a significant hazards consideration.

The NRC staff agrees with the licensee's determination and adds the comments below regarding the three standards of 10 CFR 50.92(c).

Operation of Turkey Point Units 3 and 4 in accordance with the proposed changes described above would not:

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19 (1)

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

No increase in accident probability will result from reducing reactor power sooner when core limits are exceeded, because reducing power sooner will bring the reactor back to within the required limits at an earlier time.

Also, restricting the reactor to three-loop operation removes the possibilities for thermal stresses and temperature gradients associated with asymmetric coolant flow which could accompany operation with only one or two coolant loops.

In addition, since there is no change in the design basis accidents, no increase in consequences is possible.

(2)

Create the possibility of a new or different kind of accident from any accident previously evaluated.

Because the changes merely deal with reducing the reactor power sooner to br ing the reactor back within required limits, there is no new or different kind of accident created.

The removal of alternative conditions of opera" tion; such as two-loop operation, does not create a new or different kind of accident because no new operating conditions are incorporated.

(3)

Involve a significant reduction in a margin of safety.

The proposed changes increase the margin of safety by requiring the reactor to reach Hot Standby sooner, by limiting the operation to three loops, and by providing clear guidance for timely notification of authorities.

The staff has made a final determination that the above changes do not involve a significant hazards consideration.

The staff further concludes that, throughout the RTS, similar proposed changes involving safety limits and safety settings found primarily in Part 2, which are as restrictive as, or more restrictive than, the CTS, meet the standards of 10 CFR 50.92(c) and do not involve significant hazards considerations.

~Exam 1e 2:

LCOe The CTS provided no requirements for containment air temperature.

The changes proposed in Section 3/4. 6.1. 5 of the RTS (page 3/4 6-7) include new limits on containment air temperature, an action statement that requires reactor power reduction if the limits are exceeded for a stated time interval, and stated operational mode applicability (modes 1 through 4) that applies the LCO and action statement to those operating modes in which a serious accident is most likely to occur (at power or while the reactor is hot).

In Appendix A of Attachment II of the Application, pages 3/4 6-9 and 6-lO, the licensee evaluated these additional and more limiting requirements against the three standards of 10 CFR 50.92(c) and concluded that the proposed changes do not involve significant hazards considerations.

The staff agrees with the licensee's conclusion and adds the following comments.

Because additional and more limiting requirements are imposed on plant operation, the probability of an accident and its consequences are reduced.

It is less likely that the containment and the equipment in it will fail or cause an accident due to high temperatures.

Thus, there would be no increase in probability or consequences of an accident.

There will be no new or different kind of accident created,

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20 nor any reduction in a margin of safety because the added requirements (addition of an LCO, action statement, and mode applicability) are all as restrictive, or more restrictive, than the CTS.

Therefore, the staff concludes that the three standards of 10 CFR 50.92(c) are

met, and has made a final determination that there are no significant hazards considerations.

The staff further concludes that, throughout the RTS, proposed changes consisting of additional or more restrictive LCOs, action statements and mode applicability involve no significant hazards considerations.

~Exam le 3:

Surveillances and tests In the containment air temperature example discussed

above, no surveillance or test requirements were provided in the CTS.

The proposed changes in the RTS in Part 4 (page 3/4 6-7) add a sur veil'lance requirement to determine containment temperature at various locations at least once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

In Appendix A of Attachment II of the Application, pages 3/4 6-9 and 6-10, the licensee evaluated these changes against the three standards of 10 CFR 50.92(c),

and reached the conclusion that no significant hazards considerations are involved.

The staff agrees with the licensee's conclusion and offers the following comments.

Adding new surveillance requi rements or increasing the frequency of existing surveillances for equipment that is important to safe plant operation can provide additional knowledge of the plant status so that operators can take timely cor-rective action if needed.

Therefore, this helps to reduce the probability or consequences of an accident.

Such action also helps to prevent new or different kinds of accidents from occurring, and helps to maintain desired margins of safety.

The staff concludes that the three standards of 10 CFR 50.92(c) are met, and has made a final determination that there are no significant hazards consider-ations'nvolved.

The staff also concludes that, throughout the RTS, where surveillance or tests have been added or made more extensive or frequent, there are no significant hazards considerations involved.

~Exam le e:

Administrative controls Changes proposed in Part 6 of the RTS include changes in requirements for reporting-s qualifications for operators, procedures, training scope, programs (e. g., chemistry),

records retention, high radiation areas, review and over-sight by review committees, and NRC approval of the Process Control Program.

All of, these changes are evaluated by the licensee in pages 6-1 through 6-47 of Appendix A of Attachment II of the Application.

The licensee has evaluated these types of changes against the three standards of 10 CFR 50.92(c) and con-cluded they do not involve significant hazards considerations.

The staff has reviewed all of these changes individually and agrees with the licensee's con-clusion.

Therefore, the staff has made a final determination that, the changes do not involve significant hazards considerations.

Reporting requirements are considered to be typical of administrative controls in that they do not have a strong immediate influence on the probability of an accident or its consequences (compared, for example, to the level of reactor pressure or temperature or the availability of cooling systems),

or on the kind of accident or the margins of safety.

A detailed example and analysis of the proposed changes follows.

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21 In Sections 6.9, 6.9.1, 6.9.3, and 6.9.4, beginning on page 6-15 of the CTS, requirements are described for reporting various information items to the NRC.

The RTS adds a new requirement for reporting challenges to the PORVs or safety valves.

In addition, the RTS clarifies the-reporting of changes in the analytical procedure for determining peaking factor limits.

These RTS changes do not significantly increase the probability ot consequences of an accident previously evaluated because reporting requirements are only indirectly linked to accident probability.

Adding reporting requirements in the RTS does not significantly increase accident or consequence probability, does not create a new or different kind of accident, and does not reduce safety margins.

Reporting requirements alert the NRC to the status of plant operational activities of the licensee and provide generic information for use in long-term investigative programs.

Likewise, the clarification of requirements simply removes some of the latitude for interpretation previously available to the licensee.

The three standards of 10 CFR 50.92(c) are met for these

changes, and the staff has made a final determination that there are no significant hazards considerations.

3. 4 Cate or 3 - Relaxation of CTS Re uirements Throughout Appendix A of Attachment II of the Application, the licensee has identified a number of changes characterized as "relaxations" from the CTS.

The NRC staff review has determined ther e are no cases where a significant relaxation has been made and not identified as such.

In a few cases the staff does not agree there is a relaxation, but because the licensee's characterization is conservative, it is an acceptable characterization.

Nearly all of the licensee's proposed relaxations are in Parts 3 and 4 (LCOs and surveillance requirements) of the RTS.

Only one relaxation was proposed in Part 2 (safety limits), on page 2-5 of Appendix A of Attachment II of the Application; the staff does not agr ee it is a relaxation.

Furthermore, the staff found no relaxations in Part 1 (definitions) or Part 5 (design features).

There were four relaxations proposed in Part 6 (administrative controls); these are described in Appendix A of Attachment II of the Application on page 6-13, item A. 2.c, page 6-23, items A.2.c.l and A. 2.c. 2, and page 6-26, item A.2.c.

These relaxations of administrative controls are not significant because they are minor changes which have only a weak link to operational safety.

For example, the changes proposed on page 6-23, items A. 2. c. I and A.2. c. 2 of Appendix A of Attachment II of the Application relax the time limit permitted for some reporting requirements.

Therefore, the only relaxations of any significance are in Parts 3 and 4.

Examples of proposed relaxations which are typical of those in Part 3 (LCOs) and Part 4 (surveillance requirements) and the evaluation of no significant hazards considerations are discussed below.

~fxam ie 1:

Relaxation of LCOs and surveillance frequencies In CTS sections

3. 2.4. a and b and 3.2. 5, pages

3. 2-2 and 3. 2-3, requirements on movable control rods are stated.

These include:

a limit of 0.3X potential reactivity insertion by ejection of an inoperable

rod, and a reduction of the high flux trip setpoint limit when alarms for both rod deviation and power

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22 range channel deviation are inoperable.

If either of these two alarms are inoperable, the CTS action requires that control rod positions be logged once per shift.

The licensee has evaluated a proposed relaxation of the above requirements in Appendix A of Attachment II of the Application, pages 3/4 1-33, items A. 2. c. 1,2,3, and 4, and on pages 3/4 1-34, 35, and 36.

The first (0.3X)

LCO is relaxed in the RTS by deleting the requirement.

The setpoint reduction LCO is relaxed by replacing the requirement for a setpoint limit reduction with two requirements which increase the surveillance of the two parameters.

The rod position surveillance frequency is increased from once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when the rod deviation alarm is inoperable, and the quadrant power tilt surveillance frequency is increased from once per 7 days to once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the power range deviation alarm is inoperable.

The CTS action is relaxed in the RTS by replacing the logging of rod positions once per shift with the more useful action of calculating the quadrant power tilt ratio once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The licensee also proposes to relax the surveillance frequency for rod operability from 14 days in the CTS to 31 days in the RTS.

The licensee determined that the three standards of 10 CFR 50.92(c) have been met and there are no significant hazards considerations.

The licensee's eval-uation rele'vant to the requirements of item (A)2.c discussed here is reproduced below:

"The proposed change to relax the requirements in Item 2.c above, does not involve a significant hazards consideration because these changes do not:

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

The reactivity limit in the current TS is not needed to preserve any rod ejection analysis design assumption.

Other restrictions in the proposed TS for MOVABLE CONTROL ASSEMBLIES ensure that the normal rod insertion and alignment limits are preserved thereby preserving the original safety analysis limiting assumptions related to rod position.

The current Technical Specification requirement to reduce the hi-flux trip setpoint when both the rod deviation and the power range channel deviation alarm are inoperable is replaced with more frequent surveillances.

The rod position surveillance is increased from once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when the rod deviation alarm is inoperable.

In the quadrant power tilt Technical Specification the power tilt surveillance is increased from once per 7 days to once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the power range deviation alarm is inoperable.

The increased surveillances will adequately compensate for an INOPERABLE rod position deviation or power range channel deviation alarm and are consistent with industry practice in that these are the same SURVEILLANCE REQUIREMENTS as in the Standard Technical Specifications.

Relaxing the rod OPERABILITY test surveillance from 14 to 3[1] days has no impact on control rod availability because of the insignificant.number of control rod drive failures deter-mined by the current bi-weekly surveillance test.

The proposed I

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23 surveillance reduction will also have the benefit of decreasing the likelihood of inadvertently dropping a rod during the test and reducing wear on the rod drive mechanism from the surveil-lance test.

The proposed 3[13 day test interval is also con" sistent with industry practice in that 351) days is the Standard Technical Specifications surveillance interval.

In summary, the proposed relaxations of current Technical Specification requirements do not significantly increase the probability of or consequences of a previously evaluated acci-dent because:

The 0.3X reactivity limit is not necessary to preserve any Safety, Analysis margin, setpoint reduction is not as appropriate a requir ement to compensate for an INOPERABLE rod position deviation and flux deviation alarm as the increased surveillance, the 3flg day surveillance to verify rod OPERABILITY, combined with other rod position surveillance requirements will adequately verify rod OPERABILITY.

b.

Create the possibility of a new or different kind of accident from any previously analyzed because the proposed change introduces no new mode of plant operation nor involves a

physical modification to the plant.

C.

Involve a significant reduction in a margin of safety.

As discussed in item 3a above, the 0.3X reactivity limit is not a restriction based on any safety analysis assumption, the hi-flux setpoint reduction is not necessary to compensate for any adverse impact of an INOPERABLE rod position and flux deviation alarm in the safety analysis, and the 31 day OPERABLE rod surveillance is consistent with industry practice and the Standard Technical Specifications.

Based on the above considerations, the changes included in the development of proposed Technical Specifications 3/4. 1.3. I are considered not to involve a significant hazards consideration as defined in 10 CFR 50.92."

The staff agrees with the licensee's evaluation and conclusion and adds the following comments.

In part a. there is a typographical error in that the rod operability surveillance test frequency is referred to four times as 30 days instead of the correct value of 31 days.

This example involves deleting an LCO that is not needed to preserve any safety analysis assumption.

Where this type of change occurs in the RTS, the staff concludes that there are no significant hazards considerations.

The replacement of an LCO with more restrictive sur-vei llances to accomplish the objective of monitoring plant status while reducing the risk of accident from an unnecessary plant transient is also involved.

Where

,this type of change occurs in the RTS, the staff has made a fina) determination that there are no significant hazards considerations.

This change also involves replacing an action which is not very useful (logging control rod position) with one which addresses the safety concern (power distribution) and which is explicit with its own subsection, "ACTION:".

Throughout the RTS where this type of change in action statement is made, the staff has made a final determination that there are no significant hazards considerations because the improved knowledge of plant status that results from the change does not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve. a significant reduction in a margin of safety.

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The staff also notes that this is an example where the surveillance frequency is not critical to safety because of the low rate of equipment failures, and where the test itself adds to the risk of failure.

Therefore, the staff has made a final determination that, throughout the RTS, there are no significant hazards considerations associated with relaxing surveillance frequencies of this type.

~Exam le 2:

Relaxed action statement within LCO Section 3.7.2 (pages 3.7-1 and 2) of the CTS requires that four d.c. batteries be maintained in an operable condition.

The action required, if one battery is inoperable for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , is to place both reactor units in Hot Standby (mode 3) within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The proposed change in Section 3.8.2.1 of the RTS identifies the specific four batteries, and extends the time available to be in Hot Standby to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

In Appendix A of Attachment II of, the Application, pages 3/4 8-11 through 3/4 8-14, the licensee evaluated proposed changes against the three standards of 10 CFR 50.92(c) and concluded there are no significant hazards considerations.

The relevant part of the licensee's evaluation is reproduced below.

"The proposed change to relax the action requirements to allow for a sequential unit shutdown if a battery is inoperable does not involve a significant hazards consideration because this change would not:

(a) Involve a significant increase in the probability of or consequences of an accident previously evaluated.

The proposed action statement requires within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of 'loss of a battery that one of the affected units be placed in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months followed by immediate shutdown of the other units to HOT STANDBY with the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Both units are required to be placed 'in COLD SHUTDOWN within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months following achievement of HOT STANDBY.

The proposed ACTION statement provides for a more organized method for a dual unit shutdown.

Because the proposed change would allow more preparation time to shutdown the second unit or restore the inoperable DC bus to operable status and as the likelihood of an accident being initiated during this additional short time is remote, this change would not involve a significant increase in the probability of or consequences of an accident previously evaluated.

(b) Create the possibility of a new or different kind of accident from any previously analyzed because the proposed change introduces no new mode of plant operation nor involvefs) a physical modification to the plant.

(c) Involve a significant reduction in a margin of safety because the revision allows time to recover the DC bus while shutting down one unit and preparing for an organized shutdown of the other unit.

FPL believes the advantages of these operational considerations would increase the margin of safety in the unlikely event that dual unit shutdown is required."

The staff agrees with the licensee's evaluation and conclusions and adds the following comments for clarification.

This change is proposed because it is safe~ to shut down the two Turkey Point units sequentially, rather than simul" taneously, from the control room that is common to both units.

Because it

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25 takes about 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for one Turkey Point unit to reach Hot Standby from full power in an orderly process, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is needed for two units to reach Hot Standby.

The'added orderliness of control room activities and reduced trans" ient demand on plant equipment obtained by shutting down only one unit at a time is safer than shutting down both Turkey Point units simultaneously.

Mhile this change represents a relaxed requirement on the licensee, the staff believes it is safer for Turkey Point.

Also, specific identification of the batter ies is a desired clarification.

Therefore, the staff has made a final determination that these changes do not involve significant hazards considerations.

~Exam le 3:

Relaxation of eunueil lance In the CTS, Table 4.1-1, item 18B (no page number, but follows page 4.1"1), area radiation monitors are required to be checked daily, tested monthly, and cali-brated annually.

The proposed change in the RTS would delete this as a TS requirement.

In Appendix A of Attachment II of the Application, page 3/4 3-10, item A.2.c. 1, and on pages 3'-11 through 3l4 3-13, the licensee evaluated this change against the. three standards of 10 CFR 50. 92(c) and determined there is no sig-nificant hazards consideration associated with the change.

The staff agrees with the licensee's evaluation and conclusion.

The staff notes that:

(1) these monitors are not used for automatic protection during accidents, (2) other radiation monitors provide indication of an accident by monitoring high radiation in the containment building, reactor coolant, or other process

systems, (3) these monitors will be maintained and monitored using plant procedures, and they are backed up by area radiation surveys, and (4) this change is consistent with the STS.

The staff has made a final determination that elimination of surveillances of this type, i.e,, those not required to protect the public health and safety, involve no significant hazards considerations because they do not (1) involve a significant increase in the probability or consequences of an accident previ-ously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

3.5 Cate or 4 - Relocation of Re uirements to Other Controlled Oocuments The only example of relocation of requirements from the CTS to another controlled document is described on page G-l of Appendix A of Attachment II of the Applica-tion.

Oesign requirements involving seven technical areas are proposed to be deleted from the CTS.

The licensee has evaluated these deletions against the three criteria of 10 CFR 50.92(c) on page G-4 of Appendix A.

The staff agrees with the licensee's evaluation.

The staff notes that the design requirements to be deleted from the CTS are contained in another controlled document, the Final Safety Analysis Report (FSAR).

For example, the first design feature listed on page G-1 of Appendix A of Attachment II of the Application, "Reactor Coolant System design and maximum potential seismic accelerations (Section 5.2.2 a

8 b)," is located on page 5.2-1 of the CTS and also on pages 5A-2 and 2.11-2 of the FSAR; the reference is deleted in the

RTS, The other six design features are described completely in the FSAR.

These design requirements are not used by reactor operators.

For example, the design feature cited above (Section 5.2.2

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a and b of the CTS) states the horizontal and vertical seismic acceleration limits for which the plant structures and equipment are designed.

There are no relevant operating procedures or operating requirements dealing with these seismic accelerations.

The accelerations were used to determine necessary structural strength when the plant and equipment were designed and built.

Because these design requirements are not used by the operators in their day-to-day operation of the plant, the relocation of this material is not a relaxation of. requirements.

A similar comment applies to all seven design requirements being relocated to the FSAR.

The FSAR is a better place to'locate such information.

Because there is no change to the plant or its design operating requirements, the relocation of the seven design requirements does not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

The three criteria of 10 CFR 50.92 are met, and the staff has made a final determination that there are no significant hazards considerations.

Based on the above considerations, the Commission has made a final determination that the amendment request involves no significant hazards considerations.

3.6 Public Comments No public comments were received on the notice published on May 15, 1990.

4.0 Licensin Board Comments

4. 1 Contention 6

On page 22 of a Memorandum and Order dated June 15, 1990 (which resulted from a prehearing conference on these proposed amendments),

the Atomic Safety and Licensing Board (ASLB) noted an inconsistency between the RTS and the licensee's No Significant Hazards Evaluation.

The inconsistency resulted from an error in the licensee's NSHE on page A 1"5 of Appendix A of Attachment II of the Application.

The NSHE erroneously indicated that mode reduction would not be required for 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months after inoperability had been established for a diesel generator in one train and a different component in the opposite redundant train.

This was inconsistent with the RTS, which correctly requires mode reduction. within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months (paragraph

3. 0. 3 on page 3l4 O-l).

In a submittal dated July 20, 1990, the licensee has explained and corrected the inconsistency in its NSHE.

Reference to the 14 hour1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months 's in the NSHE was deleted.

No change was necessary to the RTS itself because it was not in error.

Thus, no relaxation will occur in the time allowed for mode reduction, because the CTS also required mode reduction within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months due to CTS paragraph 3.0.1.

In addition, the ASLB urged the applicant and staff to pay detailed attention to possible risks related to continuing to operate when one power source fails.

In that regard, the staff notes that the following redundancy and diversity in electrical power supply is provided to the safety busses at Turkey Point, and that emergency procedures are in place to activate these sources as needed.

1.

Normally, the safety busses and equipment are powered by the unit auxiliary transformer receiving power from the operating unit itself.

An unrelated failure of a diesel or a startup transformer would not impact this oper ation.

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In addition, a unit startup transformer can provide offsite power from the electrical grid automatically to the safety busses and two trains of safety equipment of its associated unit, if needed.

3.

In addition, a second startup transformer (associated with the opposite unit) can be cross-connected to provide offsite power from the electrical grid to a safety bus of the unit.

This is a manual operation which the licensee has stated can be done in 20 minutes.

4.

In addition, an emergency diesel generator can provide onsite power to one train of safety equipment on the unit, independent of the three sources of power above.

This is an automatic operation.

5.

In addition, a second emergency diesel generator can provide onsite power to a second train of safety equipment on the unit, independent of the four sources of power above.

This is an automatic operation.

6.

In addition, there are five full-size cranking diesel generators, any two of which can provide power simultaneously to the non"safety "C" bus, which can be connected to a unit's safety busses to power either train of safety equipment, independent of the five sources of power above.

This is a manual operation requiring less than 30 minutes.

The "C" bus can also be powered from the offsite grid, independent of any diesels or startup transformers.

7.

The staff further notes that the above six sources of electrical power

supply, any one of which is sufficient to shut the reactor down safely, will be augmented by the licensee in the future.

The licensee plans to install two additional sources of onsite emergency power in the form of two additional full-size emergency diesel generators, each of which will automatically start and power a train of safety equipment, if needed.

Furthermore, each can be cross-connected to either of the opposite unit safety busses.

Because all of these diverse and redundant sources of power are provided, it is not considered to be a significant risk for a unit to operate for a limited period of time with one source inoperable.

4.2 Contentions ll 14 and 30 In another Memorandum and Order dated July 17, 1990, the ASLB requested the NRC staff to comment on whether three specific contentions contained serious issues.

These three contentions were numbered ll, 14, and 30 and the ASLB concerns were described in the earlier Memorandum and Order dated June 15, 1990.

The staff's discussion of these three contentions is provided below.

In addition, in the July 17, 1990 Memorandum and Order, in footnote 19 at the bottom of page ll, the ASLB indicated a concern there might be serious issues linking contentions ll, 14, and 30 to an event which occurred at the Vogtle plant on March 20, 1990 and which was reported by the NRC staff in NUREG-1410.

However, the staff regards issues identified in NUREG 1410 as generic, and actions are underway to deal with them.

The Executive Director for Operations

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28 (EDO) has requested action plans from the staff to address the issues identified, and announced his intention to monitor the resolutions of each action item.

The action plans have been completed and resolutions are underway.

The staff sees no obvious link between the NUREG 1410 issues and contentions ll, 14, and 30, and no need to revise the associated TS at this time.

However, should the resolutions of the action items necessitate changes to TS, theses will be implemented at all plants where they are needed.

A. Contention ll In its Memorandum and Order dated June l5, 1990, the ASLB noted that a concern was raised whether the applicant and staff had considered the risk associated with Mode reduction only to Mode 4 instead of Mode 5 in a specific instance involving ECCS, or more generally, Mode reduction to the first Mode beyond the LCO requirement.

The staff notes that the plant is designed to operate safely in a Mode, provided the equipment and redundancy that is needed for operation in that Mode is operable.

If it is not operable, then an action statement requires Mode reduction within a specified period of time to bring the plant into an operational Mode which is safe

and, by definition, for which the necessary equipment and redundancy is operable.

It is not necessary to shut the reactor down and defuel the core each time a component is found to be inoperable because the FSAR safety analysis has determined the combination of equipment and reactor power and coolant conditions (i.e.,

Mode) that is sufficient to make the plant safe.

The RTS reflect the FSAR and the safety analysis.

Going beyond this Mode adds an insignificant amount of safety because the plant has already been designed to accommodate accidents and transients in that Mode providing that the necessary equipment is operable.

Thus, while it is a relaxation of requirement and operational burden on the applicant to only requi re shutdown to Mode 4 in the RTS instead of Mode 5 in the CTS for the ECCS example cited in the contention, it is not a relaxation of safety.

Mode 4 is safe and the CTS were unnecessarily restrictive in this regard.

The staff further notes that the principle of reducing Mode to the first Mode beyond the LCO requirement was already established by Amendments 114 and 108 in the CTS by CTS 3. 0. l, which states in part:

"3.0:I When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action shall be initiated to place the unit in a MODE in which the specification does not apply placing it, as applicable, in:

a)

At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,"

b)

At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and" c)

At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ."

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29 Specifically with r'egard to the example involving ECCS, the RTS (and STS) would, if there were no need for ECCS in Mode 4, only require a mode reduction to Mode 4.

However, the STS and the RTS do recognize a need for ECCS in Mode 4 by the inclusion of RTS 3.5.3 applicable in Mode 4 only.

The LCO in RTS 3.5. 3 requires less than full redundant equipment on the basis, as stated in Bases 3/4. 5. 2 and 3/4. 5. 3, of stable reactivity condition of the reactor and the limited core cooling requirements.

RTS 3.5.2 requires a Mode reduction to only Mode 4, but with further degradation RTS 3.5.3 would require a reduction to Mode 5.

For this analysis, the principle of only requiring Mode reduction to the first non-applicable Mode has been maintained.

The combination of RTS 3. 5. 2 and 3. 5. 3 has been implemented on many plants on the basis of the analysis in Bases 3/4. 5. 2 and 3/4. 5. 3 and because of similarity of design it is applicable to Turkey Point.

Therefore, the ECCS example does not introduce a

serious safety issue.

B. Contention 14 In its Memorandum and Order dated June l5, 1990, the ASLB admitted Contention 14 for the purpose of addressing its concern that during the additional time in HOT STANDBY it seems to be possible to lose the ability to reduce modes and that the possible safety implications of this loss of ability require explanation.

0 The staff notes that, with loss of both of the flow paths required by RTS 3. 1.2.2, there are two additional alternate methods of adding borated water to compensate for RCS volume shrinkage while reducing the operational Mode.

In one method, Action c of RTS 3. 1.2.2 provides an identification of the Chemical and Volume Control System (CVCS) as including equipment and flow paths available for use in a mode reduction to Cold Shutdown.

As stated in Section 9.2.2 and shown in Figure 9.2-la of the FSAR the flow path from charging pumps to reactor coolant pump seal injection is an alternate to the path through the regenerative heat exchanger.

The second

method, described in the FSAR (page

3. 1.2. 6), is the injection of boric acid solution by operation of the safety injection pumps taking suction from the refueling water storage tank.

.The applicant has stated in its No Significant Hazards Evaluation (page A 3/4 1-17 of Appendix A of Attachment II to the Application submitted June 5, 1989) that 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months could be required to borate from the RWST.

The discussion below shows that at least 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months is available.

A consideration is the time to reach HOT SHUTDOWN (350~F) as determined by the volume of water in the condensate storage tank for use by the auxiliary feed water system.

RTS Bases 3/4.7. 1.3 states

that, as a minimum, sufficient water is available to maintain the Reactor Coolant System at HOT STANDBY conditions for 15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months and then cool down the Reactor Coolant System to below 350 F at which point the Residual Heat Removal (RHR) System may be placed in operation.

Below 350 F the RHR system provides the cooling function and can be aligned to take suction from the refueling water storage tank for injection of borated water.

With the loss of both of the two boration flow paths specified by RTS 3.1.2.2, Action c requires initiation of boration and going to COLD SHUTDOWN as soon as possible within the limitations of the boration and pressurizer level control of the CVCS.

As previously indicated, the reactor coolant pump seal injection would be an alternate to the path through the regenerative heat exchanger.

A second method involves the safety injection pumps taking suction from the RWST.

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30 Because the FSAR states that there are these two alternate paths for injecting an additional volume of borated water from the RWST and that the estimated maximum time is well within the available time for cool down using water from the condensate'torage tanks, the staff concludes that the ability to reduce Modes is not lost.

Therefore, there is no serious safety issue.

C. Contention 30 Contention 30 states a concern that the amendments would change the CTS at specification 3/4.4.1. 1.

The contention claims that the RTS relaxes the allowed outage time for a Reactor Coolant Loop in Mode 1 from one hour to six hours.

The ASLB observed that for this TS the loss of a coolant loop reduces heat removal capacity and that it is important that operation in this mode even for six hours be analyzed.

Ly First, the staff points out that the 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is not an allowed outage time.

Most TS Actions include a remedial action.

If the remedial action is completed within a specified time (allowed outage time), continued operation is permitted; otherwise mode reduction is required in a manner specified in the ACTION.

However, in this case RTS 3.4. l. 1 ACTION does not permit a remedial action with an opportunity for continued operation.

Rather, it consists only of a specification that requires completion of shutdown.

Throughout the STS 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months has been adopted as a standard time for achieving HOT STANDBY and has been implemented in many Westinghouse Plants.

Generally the time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to achieve HOT STANDBY allows sufficient time for the plant to be shutdown in a controlled and orderly manner, and thereby reduce the potential for challenges to safety systems and the initiation of plant transients.

The staff notes that provisions have been made for adequate heat removal capacity under all operating conditions.

If reactor power is above 45X and flow in one loop is suddenly and significantly reduced, the reactor will automatically trip.

Several diverse trip functions are provided to assure this, including overtemperature hT, loss of flow, low voltage or low frequency on pump power supply bus, and pump circuit breaker opening.

Therefore, this Action Statement will generally not be entered at power levels above 45X.

Although two-loop operation has not been fully analyzed for continuous long-term operation which is not permitted in this plant, it is clear that, because one loop can provide at least a heat removal capacity of 33 1/3X of full power, two loops'rovide adequate heat removal capacity below 45X power while shutting the plant down.

In addition, below 45X power, the overtemperature hT trip and the pump bus low voltage and low frequency trips continue to protect the plant.

The licensee's analysis (FSAR Section 7.2) shows that, for all cases in which the design limits for DNB (departure from nucleate boiling) are approached, a

reactor trip would be actuated.

Should a second loop be lost below 45X and above lOX power, the reactor would trip on loss of flow.

Many years of reactor operating experience in various transient conditions provide confidence that the plant can be safely shut down with two loops operating.

The staff concludes that provisions are in place to safely shut down the plant when one coolant loop is inoperable and that no serious issues exist.

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31 ENVIRONMENTAL CONSIDERATION 5.0 e

-.se amendments involve a change to requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements.

These amendments also involve changes in recordkeeping, reporting or administrative procedures or requirements...We have determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.

The Commission has previously issued a proposed finding that these amendments involve no significant hazards consideration and there has been no public comment on such finding.

Accordingly, these amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (10).

Pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the issuance of these amendments.

G. 0 CONCLUSION We have concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed

manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.

t Dated:

August 28, 1990 Princi al Contributors:

C.

W.

Moon A.

L. Toal ston G.

E.

Edison

I'

ML17348A534

Text

Reference:

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