ML17338A774

Safety Evaluation for Steam Generator Repair Program.Program Found Acceptable

ML17338A774
Person / Time
Site:	Turkey Point
Issue date:	05/14/1979
From:	Office of Nuclear Reactor Regulation
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ML17338A773	List: ML17338A772 ML17338A774
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NUDOCS 7906210102
Download: ML17338A774 (53)

Text

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION LICENSE NOS DPR" 31 AND DPR-41 FLORIDA POWER AND LIGHT COMPANY TURKEY POINT PLANT UNITS 3 AND 4 DOCKET NOS.

50-250 AND 50-251 MAY 1 4 1979 v 0oes1O(02

Section Tit1e TABLE OF CONTENTS

~pa e

1.0 2.0

3.0 INTRODUCTION

1.1 History of Steam Generator Operation 1.2 Reasons for Steam Generator Repair DESCRIPTION OF REPAIRED STEAM GENERATOR

2. 1 Mechanical Design and Material Changes 2.2 Heat Treatment of Tubing 2.3 ASME Code and Regulatory Guide Implementation 2.4 Removal and Reinstallation Operation 2.5 Post Installation Testing 2.6 Radiological Considerations 2.6.

1 Occupational Radiation Exposure 2.6.2 Radioactive Waste Treatment 2.6.3 Airborne Radioactive Releases 2.6.4 Liquid Waste 2.6.5 Solid Waste 2.6.6 Disposal of Steam Generator Lower Assemblies 2.7 guality Assurance EVALUATION 3.

1 Effects of. Steam Generator Design Changes 3.2 Effects of Repair Activities 3.2.

1 Protection of Safety Related Equipment 3.2.2 Other Interactions with Operating Unit 3.2.3 Fire Protection 3.3 Transient and Accident Analysis 3.3.

1 Discussion 3.3.2 Non-LOCA Accidents and Transients 3.3.3 Loss of Coolant Accidents (LOCA) 3'.4 Steam Generator Tube Rupture 3.3.5 Summary

3. 4 Radiological Consequences of Postulated Accidents 3.4.

1 Accidents During Operation with the Repaired Steam Generator 3.4.2 Accidents During the Repair Effort 3.5 Special License Conditions and Technical Specifications 3.6 Security 1-1l-l 1-3 2-1 2-1 2-2 2-2 2-2 2-3 2-4 2-5 2-11 2-11 2-12 2-13 2-14 2-16 3-1 3-1 3-2 3-2 3-4 3-7 3-8 3-8 3-11 3-13 3-14 3-14 3-14 3-14 3"15 3-17 3-17

4. 0 CONCLUSION

5.0 REFERENCES

4-1 5-1

INTRODUCTION By letter dated September 20, 1977 Florida Power and Light Company (FPL) submitted a report entitled "Steam Generator Repair Report-Turkey Point Units 3 and 4."~

This report was revised December 20,

1977, March 7, April 25, June 20, August 4, and December 15, 1978 and January 26, 1979.

The report describes a proposed program to repair the six steam generators on Units 3 and Units 4 by replacing the lower assembly, including the tube

bundles, of each generator.

Me determined that the proposed program requires our review, approval and issuance of license amendments.

Our evaluation of this program is presented in this report.

A Notice of Proposed Issuance was published on December 13, 1977 (42 F. R. 6259.)

FPL plans to repair all six steam generators in Turkey Point 3 and 4.

The unit 4 steam generators have the most tubes plugged and therefore will be repaired first.

The repair of Turkey Point 3 steam generators is expected to be started about one year later.

Since power demands in the FPL system peak in the summer, and the repair is expected to take from six to nine months per unit, the repair should be started in the fall in order to be completed before the next summer peak demand.

When FPL system submitted the repair plan on September 20, 1977 the corporate plan was to be prepared to start the repair for Unit 4 in October 1978.

The repair of unit 4 steam generator is now not expected to start before fall of 1979.

The steam generator repair program proposed by FPL for the Turkey Point Plant is similar to the one proposed by Virginia Electric Power Company (VEPCO)~,s,4 for the Surry Station (plant).

The two plants are similar.

Each of the plants contain two Mestinghouse three-loop PMR units that commenced commercial operation in 1972 and 1973.

Both plants originally used a sodium phosphate secondary water chemistry treatment and both plants changed to all volatile chemistry treatment (AVT); Turkey Point in late 1974, Surry in early 1975.

The repair program of the Surry units was approved on January 19, 1979.

Histor of Steam Generator 0 eration Turkey Point Units 3 and 4 began commercial operation on December 14,

1972, and September 9,

1973, respectively.

Like almost all units with U-tube design steam generators, they began operation using a sodium phosphate secondary water chemistry treatment.

This treatment was designed to remove precipitated or suspended solids by blowdown and was successful as a scale inhibitor.

However, during early use many PMR U-tube steam generators with Inconel 600 tubing experienced stress corrosion cracking.

The cracking was attributed to free caustic which can be formed when the Na/P04 ratio exceeds the recommended limit of 2.6.

In

I I

1-2

addition, some of the insoluble metallic phosphates, formed by the reaction of sodium phophates with the dissolved solids in the feedwater, were not adequately removed by blowdown.

The reaction products of these inpurities and of corrosion products with the sodium phosphates tended to accumulate as sludge on the tubesheet and tube supports.

In the sludge pile and associated crevices in the central region of the tube bundle where restricted water flow and high heat flux occur, the soluble sodium phosphates became concentrated by evaporative processes and precipitated.

This phophate precipitation (hideout) at crevices in areas of the steam generator, noted above, caused localized wastage resulting in thinning of the tube wall.

The problem of stress corrosion cracking was corrected by maintaining the Na/P04 ratio below 2.6.

Although the recommended Na/PO ratio was maintained, it did not correct the phosphate hideout problem $ r the wastage of the Inconel-600 which increases as the sodium/phosphate ratio is lowered.

Largely to correct the wastage and caustic stress corrosion cracking encountered with the phosphate treatment for the secondary coolant have now converted to an all volatile chemistry (AVT).

Both Turkey Point 3 and 4 were converted around August, 1974.

In 1975, radial deformation, or the so-called "denting", of steam generator tubes occured in several PMR facilities including Turkey Point 3

and 4, after 4 to 14 months operation, following the conversion from a sodium phosphate treatment to an AVT chemistry for the steam generator secondary coolant.

Tube denting is most severe in rigid regions or so-called "hard spots" in the tube support plates.

These hard spots are located in the tube lanes between the six rectangular flow slots in the support plates near the center of the tube bundle and around the peripherial locations of the support plate where the plate is wedged to the wrapper and shell.

The hard spot areas do not contain the array of water circulation holes found elsewhere in the support plates.

The phenonmenon of denting has been attributed to the accelerated corrosion of the carbon steel support plates in the annular spaces where the tubes intersect the support plate due to buildup, by processes analogous to phosphate hideout,'f an acid environment in the crevices, containing chlorides.

The resultant corrosion product (magnetite) from the carbon steel plate occupies approximately twice the volume of the material corroded.

Thus, the continuing corrosion exerts sufficient compressive forces to diametrically deform the tube and crack the tube support plate ligaments between the tube holes and water circulation holes.

As a result of the tube support plate deformation, the rectangular flow slots began to "hourglass;" i.e., the central portion of the parallel flow slot walls have moved closer so that some of the flow slots are closed or narrower in the center than at the ends.

On September 15, 1976, during normal operation, one U-tube in the inner-most parallel to the rectangular flow slots in steam generator A at Surry Unit No.

2 rapidly developed a substantial primary to secondary leak (about 80 gpm).

After removal of the damaged tube and subsequent laboratory analysis, it was established that the leak resulted from an axial crack, approximately 4-1/4 inches in length, in the U-bend apex due to intergranular stress corrosion cracking that initiated from the primary

1"3 side.

Since the initial parallel flow slot wall in the top support plate has moved closer, the support plate material around the tubes nearest this central portion of these flow slots has also moved inward, in turn forcing an inward displacement of the legs of the U-bends at these locations caused increase in the loop strain and ovality of the tubes at the U-bend apex.

It is this additional increase in strain at the apex of the U-bend which is believed to have initiated stress corrosion cracking of the Inconel 600 alloy tubing exposed to PMR reactor coolant.

Similarly, leaks have developed in severely dented tubes by primary side stress corrosion as a result of the increase in strain.

Subsequent to the above leak we imposed augmented inservice inspection requirements on Surry Units 1 and 2, Turkey Point Units 3 and 4, San Onofre Unit 1 and Indian Point Unit 2.

In addition, operating restrictions and limited periods of operation, typically six months, between inspections-are also imposed on severely degraded units, i.e.,

Surry Units 1 and 2 and Turkey Point Units 3 and 4.

The augmented inspection requirements include an assessment of the magnitude and progression of tube denting, and support plate deformation and cracking.

Reasons for Steam Generator Re lacement The six steam generators'at Turkey Point Units 3 and 4 have all undergone a significant amount of degradation since they began operation.

The wastage and denting phenomena, discussed earlier, have led to tube wall thinning, support plate flow slot hourglassing and plate ligament

cracking, tube denting, stress corrosion cracking, and several instances of reactor coolant leakage through cracked tubes.

As of May 1979, tube plugging for various reasons has resulted in removing 17.5X of the steam generator tubes in Unit 3 and 20.5X of the tubes in Unit 4 from service.

Due to the on-going denting problem, the certainty that additional tube plugging can result in power derating, and the economic considerations of operating the two units at substantially reduced

power, FPL submitted~

a.

proposal for the repair of the degraded portions of the steam generators.

2-1 2.0

2.1 DESCRIPTION

OF STEAM GENERATOR REPAIRS Mechanical Desi n and Materials Chan es During 1975 several modifications were made to the steam generators to increase the circulation ratio.

The modifications consisted of removing the downcomer resistance

plate, improving the moisture separators, modifying the blowdown arrangement inside the steam generators, installing tube lane blocking devices, and modifying the feedring.

These modifications will be retained or improved upon in the repaired steam generators under the pro-posed repair program.

Also, additional modifications, as'iscussed

below, will be incorporated.

A flow distribution baffle plate, located 18" above the tubesheet, will be used.

The baffle plate in designed to assist and direct lateral flow across the tubesheet

surface, minimize the number of tubes exposed to

sludge, and cause the sludge to deposit near the center of the tube bundle at the blowdown intake.

An improved blowdown system is to be incorporated.

The new system will use two 2-inch Schedule 40 Inconel internal blowdown pipes which will increase blowdown capacity.

The blowdown intake location is coordinated with the baffle plate design so that the maximum intake is located where the greatest amount of sludge is expected to deposit.

The repaired generators will have all the tubes expanded to the full depth of the tubesheet to eliminate the potential contaminant concentration

sites, The tube support plate material will be changed from carbon steel to SA-240 Type 405 ferritic stainless steel.

The new baffle plates will also be constructed of SA-240 Type 405.

This material is much more corrosion resistant in the chemistry expected during operation of the steam generator than in the currently used carbon steel.

Corrosion of SA-240 will result in an oxide which is protective under conditions in which carbon steel corrodes rapidly, as demonstrated by laboratory tests.

The new tube support plates will have a quatrefoil design.

The quatrefoil

design, consisting of four flow lobes and four support lands, provides support to the tube while allowing water flow around it.

The design has a

lower pressure drop across the thickness of the plate than the existing drilled circulation hole design and results in higher average flow velocities along the tubes, which should prevent sludge deposition.

Also, the tubes will be recessed slightly into the tubesheet holes and then welded to the tubesheet cladding.

This design reduces entry pressure losses and eliminates locations for possible crud buildup.

2-2

. Since the circulation ratio will be greater in the repaired generators, modifications to the moisture separator equipment will be made to accom-modate this increase, and minimize moisture and soluble corrodent species carryover into the turbines.

The new lower shell assemblies will have additional access ports that will improve the ability to inspect the tubesheet and flow distribution baffle, and will assist in sludge lancing.

A 2-inch nozzle is being added to the upper shell to facilitate the wet layup of the steam generators during periods of inactivity.

This nozzle can be used for addition of chemicals to maintain water quality.

To lessen downtime and facilitate maintenance and inspection, a 3/8-inch primary shell drain is included in the channel head of the repaired generators to improve drainage of the channel head.

Also closure rings will be welded inside the channel head at the base of each primary nozzle so that closure plates can be installed during primary chamber maintenance.

2 ~ 2 Heat Treatment of Tubin 2.3 The Inconel 600 tubing in the repaired steam generators will be thermally treated to produce a microstructure with improved resistance to stress corrosion cracking by PMR reactor coolant.

In addition, the tubes in the innermost eight rows of the bundle will be stress relieved after bending to minimize residual stresses.

Several benefits are expected to result from this reduction of residual stresses.

These include improved resistance in stress corrosion cracking in NaOH and to intergranular attack in sulphur-containing species.

ASME Code Re ulator Guide Im lementation All new component parts of the repaired steam generators will be designed and fabricated to the 1974 edition of the ASME Boiler and Pressure Vessel

Code, including all addenda through Minter, 1976.

Additionally all piping weld and preps,

welding, and nondestructive examination will be in accordance with the applicable sections of the lastest edition of the ASHE Code.

Also, applicable Regulatory Guides will be utilized as identified in the FPL Report~ (Section

2. 14).

2.4 Removal and Reinstallation 0 erations The repair will consist of replacing the lower assembly of each steam generator including the shell and the tube bundle and refurbishing and partially replacing the steam separation equipment in the upper assembly.

2-3 The old lower assembly will be removed from the containment building through the existing equipment hatch and transported to a special storage facility that will be constructed on the Turkey Point site.

The new lower assemblies will arrive at the site by barge.

They will be transferred to a wheeled transporter and hauled on the existing road to the containment building equipment hatch.

Prior to the repair work, the unit will be shut down and all systems will be placed in condition for long term layup.

The reactor vessel head will be removed for refueling.

All of the normal procedures for fuel cooling and fuel removal will be followed.

The fuel will be removed from the reactor and placed in the spent fuel storage facility.

The reactor vessel head will be replaced.

The equipment hatch will be opened and access control will be established.

The biological shield wall and a section of the operating floor concrete and structural steel will be removed to provide access to the steam generator.

Guide rails will be installed for transporting the lower assembly through the equipment hatch.

After this preparatory work, the cutting of system piping can begin.

This will. include cutting and removal of sections of steam lines, feedwater lines, reactor coolant inlet and outlet lines, and miscellaneous smaller lines for the service air and water and the instrumentation system.

The steam generator supports will be disassembled and the steam generator lower assembly will be lowered and placed in a horizontal position on a transport mechanism.

This mechanism will carry the assembly through the equipment hatch.

A mobile crane will lift the lower assembly onto a transporter that will carry it to the steam generator storage facility on the site.

After removal and storage of all three steam generator lower assemblies, their replacements will be transported from the barge dock or temporary storage location to the equipment hatch.

The same machinery used to remove the lower assemblies will be used to install the new assemblies in their cubicles.

The steam generator support system will be reinstalled and the upper assembly with its refurbished internals will be mounted on the lower assembly.

After welding the two assemblies

together, the piping will be replaced and the biological shield and internal structures will be reconstructed.

Mhile the pre-operational and startup test program following these major repair activities are still being developed there will be cleaning, hydrostatic testing, baseline inservice inspections, and pre-operational testing of instruments, components and systems.

Then the reactor will be refueled and startup tests will be performed.

The performance of the repaired steam generators will be tested for moisture carryover and verification of thermal and hydraulic characteristics.

2.5 Post Installation Testin A detailed preoperational testing program will be carried out by FPL prior to fuel loading to reestablish the integrity of the reactor coolant system and the main steam and feedwater

system, to ensure that all systems are in

operating condition and to provide baseline data f'r future performance evaluation.

Hydrostatic pressure tests will be performed as well as the baseline inservice inspection of piping.

The fuel manipulator crane will be re-assembled and tested.

After the residual heat removal system has been tested and placed in

service, fuel will be transferred to the reactor vessel.

One third of the fuel assemblies placed in the vessel will be new fuel assemblies and the operation will not differ significantly from a normal refueling.

During the initial startup of the unit, tests will be performed to verify the thermal and hydraulic performance of the nuclear steam supply system.

FPL has not yet completed the preparation of detailed procedures for preoperational testing and startup of the unit after completion of the steam generator repairs.

We will review the detailed procedures prior to fuel loading to verify that adequate testing will be performed to ensure safe startup of the unit after completion of these repairs.

Radiolo ical Considerations A major aspect of the repair effort is its radiological impact, including the occupational exposure accumulated during the repair effort and the radiological effluents released from the site.

These considerations are discussed below.

Battelle-Pacific Northwest Laboratories (PNL) has performed a generic radiological assessment of steam generator repair and disposal under contract to the NRC, which has been published in a separate NRC report, NUREG/CR-0199, "Radiological Assessment of Steam Generator Removal and Replacement."

The PNL estimates of occupational exposures (man-rems) developed in this report were derived by dividing the repair program into sub-activities

("maintenance activities") and determining the estimated exposure rate for each sub-activity.

The sub-activity man-hours multiplied by the corresponding exposure rates in rem per hour gave the exposure in man-rem for each sub-activity.

The total exposure for the repair program is the sum of the exposures for each of the sub-activities.

Repair program sub-activities were defined by PNL from a composite of the work descriptions for the repair of the steam generators at Surry and Turkey Point as determined by VEPCO and FPL.

Man-hour estimates for each sub-activity were developed by PNL based on prior experience with similar activities

,and on standard estimating techniques.

Exposure rates were based on information from several sources including data from measurements made at several operating PWRs including the Turkey Point Units.

PNL usually selected exposure rate values on the high end of the range of values measured at the several plants.

The PNL estimates of occupational exposures are intended to be conservative and represent upper bound values.

The PNL estimates are presented as a range of values.

The

2"5 PNL lower value was estimated assuming credit for various techniques to reduce exposures, e.g., providing water shielding by maintaining high steam-generator water levels, remote tooling and distance where applic-able.

FPL has committed to these types of techniques, consequently, it is appropriate to compare the PNL lower value with the FPL estimates.

The FPL occupational exposure estimates include a detailed estimate of doses based on major job functions of 1300 man-rem per Unit.

These detailed estimates do not include dose savings from use of temporary shielding and local decontamination or dose costs from implementation of these.

However, FPL has estimated a range of doses for the steam generator repair program of from 650-1450 man-rem per Unit.

The range of doses presented represents the best FPL judgment with respect to the. predicted worker doses considering uncertainties in prediction of job man-hours and radiation fields.

The radiation field uncertainties consider the effective-ness of temporary shielding and the time required to place such shielding.

Therefore, although FPL has not included the effect of temporary shielding and local decontamination in its detailed estimate, FPL has considered it in its predicted range of doses.

For comparison purposes in this report, we are evaluating the PNL lower estimate (3380 man-rems)s versus the FPL detailed estimate.

The FPL estimates are generally lower than the PNL lower values because the actual plant data are lower than the PNL radiation field estimates.

The FPL dose estimates are based on a range of'adiation field values from actual in-plant surveys at Turkey Point.

The estimates assume occupancy is in an average radiation field.

FPL has. stated that use of temporary shielding will be determined based on radiation surveys and an estimate of the dose savings from use of shielding compared to the dose incurred from installation of the shielding.

Me expect the actual radiation fields to be within the range of values given in the report.~

Based on our evaluation of FPL and PNL assumptions, as discussed in the following paragraphs, we have concluded that the FPL estimate should be more representative of the actual doses.

Me have included the PNL estimate for comparison purposes.

The FPL estimates include 200 man-rems for miscellaneous activities such as supervision, quality assurance and health physics.

Me have divided the estimate equally between the removal and re-installation phase in this evaluation to permit comparison with the PNL estimates.

PNL also provides estimates of radioactive effluents which could be released as a result of the repair effort.

The estimates given in this report are on a per Unit basis, i.e., repair of 3 steam generators, unless otherwise noted.

2. 6.1 Occu ational Radiation Ex osure Separation, disassembly, removal and re-installation of the repaired steam generators must be done in radiation fields.

Federal regulations as specified in 10 CFR Section

20. 1(c), state that licensees should make

2-6 "every reasonable effort to maintain radiation exposures...

as low as is reasonably achievable" (ALARA).

The FPL efforts to reduce occupational exposures to ALARA levels are addressed in this section.

The repair program activities can be broken down into four major categories:

post-shutdown preparation, steam generator removal, installation of the repaired steam generators, and disposal of portions not reused in the re-paired steam generators.

All of the activities associated with the repair activities and return to power have been incorporated into the dose estimates.

These include health physics and quality assurance/quality control activities.

2.6. 1.1 Post Shutdown Pre aration The post-shutdown activities include defueling the reactor and storing the spent fuel in the storage pool.

The defueling activities will be similar to a normal refueling except that the entire core will be unloaded and the reactor vessel head reinstalled.

The time involved in defueling an entire core will be similar to the time involved in defueling, shuffling and refueling 1/3 of a core.

Since the radiation fields will be essentially the same as for a normal refueling, consequently, the expected occupa-tional exposure should be similar to a normal refueling.

Following defueling and prior to starting removal of the first steam generator, temporary structures will be installed to facilitate the steam generator separation and removal activities.

Mater will be kept in the steam generators for shielding value as long as practical (until the pipes are to be cut).

These structures include contamination control envelopes around the reactor coolant piping at the separation

points, temporary ventilation systems, scaffolding, and construction of a temporary platform with guide rails at the equipment hatch to facilitate removal of the steam generator lower assemblies.

The preparation activities also include radiation surveys and local decon-tamination.

Portions of the concrete shield walls will be removed to per-mit later removal of the steam generator lower assemblies.

Some small sections of containment internal structures must also be removed to permit removal of the lower assemblies.

The thermal insulation around the steam generator, reactor coolant and main steam piping will also be removed.

A new 250 ton construction hoist will be placed on the polar crane bridge because the existing trolley is not capable of handling the lower assemblies.

Load testing of the new hoist will assure that current OSHA safety standards are met.

In order to reduce occupational exposures many of the activities will be performed with the steam generator secondary side partially filled with water to lower radiation fields.

FPL has estimated a total dose of 257

2-7 man-rem per Unit (excluding refueling) for these post-shutdown preparation activities.

The major portion of this dose estimate is attributed to instal-lation of. temporary structures, local decontamination and removal of insulation.

FPL has not provided a detailed estimate for installation of temporary shielding.

FPL states that the need for temporary shielding will be treated on an individual case basis.

The need for shielding will be evaluated based on the dose savings for performing the job with shielding versus the dose incurred during installation and removal of the shielding.

PNL (NUREG/CR-0199)s has estimated an occupational dose of 450 man-rem for the post-shutdown preparation activities including 20 man-rem for defueling.

The PNL estimate also assumes control of-the steam generator secondary side water level to shield radiation emanating from the primary side corrosion products.

PNL has included an estimate of 144 man-rem for radiation sur-

veys, local decontamination and installation of shielding.

It is our opinion that some shielding and local decontamination will be necessary.

As discus-sed above, although FPL has not provided an estimate for installation of shielding in the detailed dose estimate, the range of dose estimates FPL has provided does consider the effectiveness of temporary shielding and the time required for installation of the shielding, and is based on FPL knowledge of plant specific design.

2.6.1. 2 Steam Generator Removal Removal activities include removal of the main steam lines, feedwater lines, reactor coolant inlet and outlet and miscellaneous pipe segments.

These must all be removed to provide clearances in the steam generator area.

The highest exposures will most likely occur during preparation and cutting of the reactor coolant piping and cutting and removal of the steam generator upper internals because of the manhours required in the radiation areas to complete the cutting.

The reactor coolant system pipe cuts will be performed in a contamination control envelope with a ventilation system containing a

HEPA filter to minimize the spread of airborne particulates.

FPL plans to use remote cutting tools wherever possible to minimize the time personnel stay in radiation areas.

It is planned to used mockups to familiarize personnel in the specifics of the cutting operations including space restraints, protective clothing, and special tasks required.

The familiarization training should minimize time spent in radiation fields.

The cut reactor coolant pipe ends, in addition to being sealed may be shielded to reduce radiation streaming from the internal surfaces.

The steam generator upper shell will be cut and removed from the lower assembly and stored on the containment operating floor.

Remote cutting tools will be used wherever possible.

The expected low contamination levels on the secondary side preclude the necessity of using contamination control envelopes at this location to control the spread of airborne activity.

The steam generator wrapper and upper internals will be cut from outsi'de the steam generator after the upper shell has been removed.

The steam generator water level will be kept high to shield personnel making the cuts from radiation emanating from the lower shell internals.

The PNL dose estimate for cutting the wrapper assumed the cut would be

2-8 performed from inside the steam generator upper shell in much higher radiation fields because PNL assumed no credit for shielding from keeping the water level high.

The-FPL estimate of occupational exposure to cut the wrapper is lower than the PNL estimate because it is based on radiation fields actually measured at Turkey Point which are lower than those assumed by PNL.

All openings in the steam generator lower shell will be sealed with welded metal seals prior to removal of the steam generator lower assembly from the containment.

The sealed assembly will be rigged for lifting, its supports will be disassembled, and it will then be removed from the containment.

The upper shell and most of the internal moisture separation equipment will be reused.

The upper shell will be prepared for reinstallation on the new steam generator lower assembly.

The contribution to the occupa-tional exposures will be minimal due to the low contamination levels expected on secondary side portions of the steam generator and the ambient radiation levels at the work areas.

All three existing generators will be removed before any of the new generator sections are brought into the containment.

FPL has estimated a

total occupational exposure of 436 man-rem per Unit for the removal activities.

PNL (NUREG/CR-0199) has estimated a dose of 1100 man-rem for the removal phase.

2.6.1.3 Installation of Re aired Steam Generators The installation phase involves bringing in and installing the new lower shell assemblies, attaching the upper shells, transporting and reinstalling all the removed. piping and associated transition pieces, reconstructing the concrete walls removed earlier, removing all temporary work structures,

cleanup, performing preoperational structural integrity tests, refueling and preparing the containment for startup tests prior to return to power.

Similar to the removal situation and for the same

reasons, the major dose contribution to the installation activities is expected to be from recon-necting the reactor coolant system piping.

To minimize radiation exposure, an automatic welding device will be used.

PNL (NUREG/CR-0199)s has estimated a savings of 500 man-rem per generator (1500 man-rem per Unit) based on using remote welding as compared to manual welding.

This yields a.total PNL estimated exposure of 1800 man-rem per unit for the instal-lation phase.

FPL has estimated the exposure for this phase to be 569 man-rem per Unit.

The PNL estimate assumed worker occupancy in higher radiation fields than those based on Turkey Point plant surveys by FPL.

2.6. 1.4 Ois osal of Portions Not Reused Disposal also affects the occupational exposures.

This entails transporta-tion to and placement in the storage facility.

A description of this facility is contained in Section 2.6.6.

FPL has estimated 39 man-rem per Unit will be expended for the onsite storage.

PNL (NUREG/CR-0199)s has estimated 30 man-rem per Unit.

These estimates are essentially the same.

2-9 2.6.1.5 ALARA Considerations FPL has estimated 1300 man-rem per Unit will be expended for the repair program.

This estimate is based on dose rate survey data from the Turkey Point Units (see figure 3.2 of the EIA~4) estimates of man hours involved for the individual procedures and estimated savings from dose rate reduction techniques as addressed previously.

In addition, FPL estimated a range of exposures from 650-1450 man-rem/Unit based on uncertainties regarding job man-hours, radiation fields and the effectiveness of temporary shielding.

PNL (NUREG/CR-0199)s has estimated a total dose of 3380 man-rem per Unit for the whole repair program.

FPL has committed to making every reasonable effort to keep radiation exposures ALARA in accordance with 10 CFR Section

20. 1(c).

The radiation protection program followed during the repair effort will be in accordance with the FPL Health Physics Manual and its implementing procedures.

The FPL plant procedures contain sections relating specifically to health

physics, including such items as protective clothing, personnel monitoring, radiation surveys, use of temporary shielding and treatment of contaminated personnel.

FPL has stated that the Health Physics Manual reflects a manage-ment commitment to maintain occupational exposures ALARA and that the plant Health Physics Supervisor is responsible for ensuring that the ALARA policy is implemented.

FPL has stated that additional facilities will be provided for the repair effort, including a radiological protection training facility and an addi-tional health physics area.

All craft personnel will be required to take training in radiological pro-tection.

The course will include instructions and demonstration in radia-tion protection principles, theory and practice, emergency planning and the FPL Radiological Protection Program.

Personnel will be required to pass a comprehensive examination to have unescorted access in the radiation controlled area.

Those failing to pass the exam or those who take only a short basic course will need an escort in the controlled area.

Extensive training in other areas will be used throughout the repair.

FPL has stated that scale models will be used to familiarize supervisory and key craft personnel with the repair effort.

The models will be used to develop construction work plans to establish the most efficient work pro-cedures.

The models will also supplement work plans and allow supervisors and craft personnel to achieve the most efficient use of manpower which will reduce occupancy in radiation fields and, thus, reduce the total occupational dose.

These models include a model of the entire containment which wi 11 be used in conjunction with radiation survey data to establish temporary shielding requirements.

The model will also be useful in making man-rem assessments for particular work activities in radiation fields.

2-10 Other models include a scale model of the steam generator internal details and a model of laydown space requirements inside containment.

Me have con-cluded that use of the models will be a helpful tool in planning an efficient repair program and will serve to reduce occupational exposures by reducing potential occupany in radiation fields to a minimum.

FPL has stated that full scale mockups will be used to train craft personnel in steam generator cutting and welding operations.

This training will minimize occupational exposures by familiarizing personnel with the operations which should reduce the time spent for the actual operation.

FPL has stated that use of temporary shielding and local decontamination will be evaluated on an individual job basis.

The man-rem expenditure for installing and removing shielding will be compared to the man-rem savings of using the shielding.

Low radiation background areas will be established inside the containment.

Personnel needed inside containment but not immediately engaged i4 an activity will be required to wait in these areas in order to keep their exposures ALARA.

FPL has stated that special tools such as remote equipment for cutting and welding will be used whenever possible.

Use of remote and automatic tooling will save. exposure by reducing personnel man-hours to perform the job, allowing personnel to keep away from high radiation sources and allowing personnel to remain behind shielding to keep their exposures low.

Decontamination can be an effective dose reduction technique where radia-tion fields can be significantly reduced.

However, several factors must be considered when decontamination is being considered.

Chemical compati-bility of the decontamination fluid with the materials of the installed system must be proven.

Additional exposure would result from installation and operation'f decontamination equipment and processing of the radioactive waste generated.

Based on present limited experience in large scale, high volume chemical decontamination of reactor coolant systems, we believe that considerable economic impact, e.g.,

increased reactor outage time and development of equipment and procedures, would 'result from the use of chemical decontamination.

Also, the research necessary to prove the safety of such operations could have a major schedule impact.

Because of these considerations, we conclude that-chemical decontamination of the tubes is not a viable option for this program at this time.

Local work area, surfaces,

however, can be decontaminated using mild solutions.

This should provide worthwhile radiation exposure reductions for several of these areas.

FPL will evaluate the use of. local decontamination wherever dose reduction benefit can be gained similar to the evaluation for use of temporary shielding.

~Summar Me have reviewed. the

FPL, submittal regarding occupational exposures and we conclude that the repair program can be accomplished without exceeding

2"11 2.6.2 the requirements of 10 CFR 20 and that the efforts proposed to maintain occupational exposures ALARA are acceptable.

Radioactive Waste Treatment 2.6.3 Radioactive waste treatment will be used to provide treatment of radio-activity generated as a result of the repair effort so that radioactive releases to the environment are kept to a minimum.

The currently installed station waste treatment systems and temporary systems as discussed below will be used to process airborne and liquid wastes.

Airborne Radioactive Releases The Unit will be shutdown and the core unloaded before repair work is started; therefore, no gaseous wastes will be generated from reactor operations during the repair period which is expected to last about six months.

The major source of airborne radioactivity generation associated with the repair program will come from activities such as cutting and weld preparation work on open radioactive coolant piping and concrete removal.

The major source of radioactivity is expected to be particulates generated from cutting the reactor coolant system (RCS) piping.

These cuts are expected to be per-formed in a local contamination control envelope which will be ventilated to the containment through a local high efficiency particulate air (HEPA) filter.

The secondary system piping cuts and concrete removal will not require local contamination control envelopes because of the low con-tamination levels in the secondary side piping and on the concrete.

All containment releases will be exhausted by the purge system via the plant vent.

Releases will be monitored by the existing sample station and monitor on the plant vent.

There will be a slight negative pressure on the con-tainment to prevent release through the access hatches.

FPL has estimated that a maximum of 1. 1 x 10-~ Ci of air born'e radioactivity per Unit will be rel'eased to the environment as a result of the RCS piping cuts based on expected contamination levels on the reactor coolant side surfaces and expected cutting kerfs.

This activity is expected to pass through local HEPA filters to the containment atmosphere and then through the containment purge exhaust system to the environment.

Although the HEPA filters will be purchased to a removal efficiency of 99.97K, a filter efficiency of 99K was assumed for the filters.

We have inde-pendently estimated 0.27 Ci may be generated locally by cutting of the RCS piping resulting in a release of 2.7 x 10-s Ci to the environment assuming a 99K efficiency for removal of particulates by the local HEPA filter.

The difference between FPL's estimate and our estimate is due to the assumption of a different size cutting kerf.

Our estimates are based on the information given by PNL in NUREG/CR-0199 In addition, PNL has esti-mated that'. 1 x 10-s Curies may be released from secondary system piping cuts.

We, therefore, estimate the total release for pipe cutting for removal of three steam generators to be 1. 1-x 10-~ Curies.

These projected releases are less than the actual average airborne radioactivity releases during 1976 and 1977.

For 1976~~ these releases were 3.8 x 10 2 Ci of par-ticulates and 0.3 Ci of halogens (0.338 Ci combined).

During 197718, the

2-12

2. 6.4 particulate and halogen airborne releases per unit totaled

0. 726 Ci (2. 6 x 10-~ Ci of particulate activity and 0.7 Ci of halogens).

The estimated gaseous radioactive effluent per unit resulting from the repair effort, 1.

1 x 10-~ Ci of particulates, is small compared to Turkey Point historical data.

The projected airborne releases from the steam generator are expected to be well below the plant radiological effluent Technical Specifications.

FPL has submitted information to show con-formance with the design objectives of Appendix I to 10 CFR Part 50.

Although we have not completed the evaluation of this information, comparisons of the FPL data with the evaluation given in the Final Environmental Statement (FES)e for Turkey Point indicate that the steam generator repair doses will be less than the Appendix I design objectives.

The FES doses are based on total iodine and particulate releases of 0.8 Curies per year and on over 3600 Ci of noble gases per year, which are much greater than the projected releases from the repair effort.

There-

fore, we conclude that the releases will be within the Appendix I to 10 CFR Part 50 Design Objective and will be ALARA.

Li uid Waste During the steam generator repair outage, radioactive liquid waste may be generated from (1) disposal of reactor coolant water, (2) disposal of secondary coolant water, (3) local decontamination solutions. and (4) laundry waste water.

FPL is planning to store the reactor coolant for reuse after the repair is complete.

Therefore, there should be no release to the environment from reactor coolant'.

However, FPL has estimated the liquid effluent dose if the coolant were to be discharged.

The reactor coolant would be treated by the chemical and volume control system prior to any release to the environment and FPL has estimated that the resultant effluent would contain a maximum of 0.08 Ci of mixed fission and actuation products.

FPL has stated that if reactor coolant water is discharged it can be proc-cessed through a mixed bed demineralizer and the boric acid evaporator.

Based on the reactor coolant system activities given in Table 2-2 and the decontamination factors given in Table 1-3, both from NUREG-0017 (PWR-GALE Code) we have estimated the release to the environment of 2 x 10-Ci from discharging the reactor coolant system.

Actual releases will depend upon coolant concentrations at the time of processing and on the processing equipment used.

The plant liquid effluent Technical Specification must be met during the repair effort.

Secondary coolant water may be contaminated if the Unit operates with a steam generator tube leak immediately prior to shutdown.

We do not discount this possibility.

However, based on experience with previous leaks, if such a leak were present the activity levels are expected to be relatively low and would not contribute significantly to the total activity released from the plant.

2-13 Local decontamination will be used to lower radiation levels in the plant.

FPL has stated that decontamination wastes are expected to be minimal and will be treated as part of the normal liquid radwaste processing stream.

Mastes will be collected and sampled and processed or discharged as dictated by the plant Technical Specifications.

The major volume of liquid radioactive effluent releases will be from laundry waste water.

The FPL maximum estimates are based on 22,000 gallons per day being generated and released for a total of 6.6 x 10 gallons during a 300 day outage.

The waste water is expected to be of low specific activity and should not require processing before release.

However, it must be sampled to verify it is low in radioactivity concentration.

If radioactivity levels exceed those allowed by the Technical Specifications, the waste water will be processed to acceptable levels prior to release.

FPL has estimated the maximum expected release to the environment from 1'aundry wastes to be 0.47 Ci per Unit with Co-60 making up 27 percent of the total activity and Co-58 making up 36 percent of the total activity.

FPL has estimated that only 10,000 gallons per day will be released, yielding an estimated total activity release of only 0.20 Ci per Unit from this source.

Using the figure 6.6 x 10'allons, FPL has estimated a total maximum liquid release of 0.55 Ci of radio-activity (except tritium) for the repair effort for one Unit.

Me have independently estimated the total liquid release from laundry and general decontamination wastes to be 2.4 Ci.

Our estimate is based on the radioactivity releases given in Table 2-20 of NUREG-0017~ adjusted for the FPL maximum estimated release volume.

For comparison, the annual average Turkey Point release of mixed fission (not including dissolved noble gases) and ac'tivation products was 4.3 Ci of radioactivity in 1.7 x 10~ gallons per Unit in 1976 and 4.5 Ci in 1.3 x 10 gallons per Unit in 1977.

Any liquid effluent containing radioactivity would be discharged into the condenser cooling water and subsequently be discharged into the closed cycle cooling canal.

Pursuant to a Final Judgement dated September 20, 1971 in the U.S. Distric Court for the Southern District of Florida (Civil Action No. 70-328-CA; reproduced in Appendix C of the FESe) Florida Power and Light Company shall not discharge into Biscayne Bay or Card Sound any water used for cooling i'ts condensers at its generating facilities at Turkey Point.

The estimated plant liquid effluent radioactivity resulting from the repair effort is small compared to Turkey Point historical data.

The plant Technical Specifications limit the radioactivity in liquid effluents from Turkey Point Units 3 and 4 combined to 20 Ci per calendar quarter (excluding tritium and dissolved gases).

Consequently, the projected releases due to the repair program (2.4 Ci) are expected to be well within the plant Technical Specification limits.

FPL has submitted information to show conformance with the design objectives of Appendix I to 10 CFR Part 50.

Me have not completed our evaluation of the Appendix I information at this time, however, based on the results of our review-to date, we expect that the current Technical Specification limits on liquid effluents will not be reduced as

2-14 2.6.5 a result of our review of the Appendix I evaluation.

On this basis we con-clude that the Technical Specification limits will assure that releases from the steam generator repair activities will be well within the Appendix I design objectives for liquid effluents.

Solid Maste Radioactive solid wastes generated during the repair effort will include contaminated building materials used to construct temporary structures, concrete removed during the repair, miscellaneous piping, disposable pro-tective clothing, solidified liquid wastes, the lower sections of the steam generators and portions of the upper internals not reused.

-The disposal of the lower'ections of the steam generators is discussed in Section 2'.6.

The building materials used in temporary work structures should be free of any significant contamination.

Only those materials expected to be used for a temporary contamination envelope around the reactor coolant piping would be exposed to significant contamination from airborne particulates resulting from the cutting operations.

The other structures will be exposed to such contamination as may result from cutting the secondary piping.

The secondary system contamination levels are very small and cutting will not generate significant contaminants.

To facilitate the steam generator lower assembly removal some concrete will be removed from the biological shield surrounding the steam gene r ators and from other structures.

FPL has estimated a total of 1600 cubic feet (about 45.4 cubic meters) of concrete will be removed per Unit with a total activity of 3. 1 pCi.

The PNL estimate (NUREG/CR-0199)s agrees with the FPL estimate.

A major portion of the volume of solid radioactive waste generated (other than the lower sections of the steam generators) will be rags,

trash, disposable protective clothing and miscellaneous tools and building materials.

FPL has estimated about 25,800 cubic feet (about 730 cubic meters) of such waste containing approximately 100 Ci of radioactivity will be packaged and shipped to a burial facility.

In addition, FPI has estimated 30 Ci of activity will be contained in evaporator bottoms and spent resins.

FPL has estimated the repair of one Unit will result in a total solid waste volume of 27,400 cubic feet (780 cubic meters) containing 130 Ci being shipped to a licensed burial facility.

The FPL estimates are based on typical quantities and types of wastes generated during a normal refueling outage.

PNL (NUREG/CR-0199)

'has estimated a total of 81,000 cubic feet (2,290 cubic meters) of solid radwaste will be generated during the repair of one Unit.

This compares with the average amount of radio-active solid waste shipped (per Unit) of 25,400 cubic feet (720 cubic meters) and 240 Ci during 1976~~

and 19,000 cubic feet (539 cubic meters) and 210 Ci during 1977.

All radioactive waste shipments will conform to NRC and Department of Transportation (OOT) regulations.

2-15 2.6.6 Ois osal of Steam Generator Lower Assemblies The steam generator lower assemblies will comprise the largest source of radioactive waste requiring disposal.

Several options for the disposal of the lower assemblies were considered:

(1)

Immediate intact shipment to a licensed burial facility; (2)

Immediate cut-up and shipment to a licensed burial facility; (3)

Onsite storage until facility decommissioning.

Because of the size and packaging involved, the only method for immediately shipping the assemblies intact would be by barge.

At present, there are no licensed burial facilities with receipt capabilities available.

Therefore, this option is not viable for the immediate disposition but may become an option in the future.

Immediate cut-up and shipment is possible now with transportation by truck or rail.

The assemblies could be cut into suitable sized segments and packaged and transported.

Cutting of the assemblies and subsequent handl-ing would result in increased occupational exposures due to the activity on the surfaces exposed to reactor coolant.

Some dose reduction could be achieved by decontamination of the reactor coolant surfaces.

However, effective decontamination factors may not be achievable due to presence of a significant number of plugged tubes which would prevent decontamination chemicals from entering approximately 19K of the tubes.

Reduced exposures due to decontamination would be accompanied by a signif-icant increase in decontamination solution liquid radioactive wastes.

These wastes would have to be processed and solidified.

PNL (NUREG/CR-0199) has estimated a total exposure of 810 man-rem for immediate cut-up and ship-ment following chemical decontamination.

We conclude that immediate cut up and offsite shipment will cause an unnecessary man-rem burden on the workers without providing a significant operational benefit to FPL and to the public as compared to onsite storage as discussed below.

FPL has proposed long term onsite storage to allow for decay of radio-activity to relatively low levels to minimize radiation exposures before processing for shipment.

The lower assemblies would be stored in an engineered storage facility specifically contructed for this purpose.

Such storage would provide for FPL responsibility and control of access and exposure to the assemblies until the radiation has decayed to levels that will allow easy disposal (e.g., Unit decommissioning).

Based on decay of the expected radioactive corrosion products it is estimated that storage for 30 years can reduce the radiation levels to less than 1X of those expected when the assemblies are removed from containment.

The assemblies will be sealed with steel plates or plugs prior to removal from containment to eliminate airborne particulates from being released from internal surfaces.

Internal decontamination will not be necessary because of the seals.

Some surface contamination will be present on the outside

2"16 of the assemblies.

FPL has stated that the external surfaces will be decon-taminated such that removable contamination levels will be less than 2200 dpm/100 square cm prior to removal from containment.

Therefore, any release to the environment from transport of the assemblies to the onsite storage facility should be negligible.

The onsite storage facility will be a concrete structure approximately 110 ft x 60 ft with a height of 17 ft.

The outside walls will be approximately 2 ft thick.

The facility floor is earthen with no provisions for collecting water.

No water accumulation is expected since the roof is watertight and the generators will be drained prior to storage; Because the external con-tamination levels will be <2200 dpm/100 square cm airborne releases from the external surfaces of the generators are not expected.

FPL has proposed quarterly surveillance of the facility consisting of visual inspections and random swipes of the generators and area radiation surveys to assure that no airborne contaminants are being released from the facility.

There will be a limited amount of direct radiation which penetrates the storage building walls.

Based on the maximum expected radioactive inventory of the steam generators and the shielding of the storage facility FPL has estimated, using commonly accepted practices, an annual dose of less than one mrem to an individual at the site boundary.

We have reviewed the bases for this estimate and consider the bases acceptable.

We conclude that the expected radiation levels on contact with the outside of the facility walls are approximately the levels for unrestricted areas specified in 10 CFR Section 20.105.

If upon completion of the storage phase FPL finds levels in excess of 10 CFR Section

20. 105 FPL will be required to provide adequate control and posting pursuant to 10 CFR Section 20.203.

We have reviewed the FPL proposed surveillance program for the storage facility and find it acceptable.

We conclude that the program will provide adequate assurance that there will be no significant releases from the storage facility.

The use of an onsite storage facility will minimize immediate occupational exposures since no immediate disassembly and packaging for equipment is necessary.

In addition, the long storage time will allow for significant decay of radioactivity so that ultimate disposal at the end of station life will not be a significant occupational dose impact.

Therefore, we conclude that use of an onsite storage facility is in accordance with ALARA philosophy.

We have reviewed the FPL proposed method of storage and conclude that there is reasonable assurance that this storage will not endanger the health and safety of the public and is acceptable.

In addition, we conclude that the measures to be taken to control and monitor this storage will keep occupational exposures and radioactive effluents as low as reasonably achievable.

2-17 2.7

~Summar Me have concluded that FPL's efforts to maintain occupational exposure to ALARA values during the repair effort are reasonable and adequate radiation protection will be achieved.

We have further concluded that the radioactive effluents which may be released as a result of the repair effort are less than those expected during normal operations, can be maintained within the radiological effluent Technical Specifications and will not affect the health and safety of the public.

ualit Assurance The quality Assurance Program for the repair of the. steam generators will be in accordance with the Florida Power and Light "FPL guality Assurance Topical Report",

(FPLT(AR 1-76A),

except as amplified in Section 3.6. 1 of Rev.

3 of the FPL Steam Generator Repair Report.

Me find these amplifica-tions to be acceptable clarifications of FPL commitments contained in FPLT(AR 1-76A, Rev.

2.

Work performed by Bechtel on the repair of the steam generators will comply with the "Bechtel guality Assurance Program for Nuclear Power P 1 ants, " (Bg-TOP-1). 9 The guality Assurance Program for the design and fabrication of the steam generator replacement lower shell assemblies and other components will be in accordance with the Mestinghouse Electric Corporation topical report (WCAP-8370 Rev.

SA),

Each of the above reports has been reviewed by the NRC for compliance with Appendix 8 to 10 CFR Part 50 and has been found acceptable.

Me now have reviewed the aforementioned reports with specific consideration for the proposed steam generator repair.

Based on our review we find that:

(1) the repair activity is within the scope of the approved

programs, and (2) adequate controls exist within the approved programs for the proposed work activities.

Accordingly, we find the provisions established for the quality related activities associated with the repair of the steam generators acceptable.

8

3"1 3.0 EVALUATION Several design

changes, as discussed above, will be incorporated in the repaired steam generators.

Our evaluation of these changes is given below.

3.1 Effects of Steam Generator Desi n Chan es The existing steam generators contain large amounts of sludge which has contributed to their previously discussed degradation.

Since an AVT secondary water chemistry treatment will be used when the repaired steam generators begin operation, and residual phosphates will not be present in the system, any sludge which accumulates should not be of a chemical com-position that could lead to degradation of the repaired generators.

Along with the absence of phosphates, planned condenser retubing and the installation and use of condensate polishers will essentially eliminate sludge.

Furthermore, even if sludge should form, we concur that a flow distribution baffle plate should minimize, or at least reduce, the number of tubes exposed to the sludge, and cause the sludge to deposit near the blowdown intake.

Use of this baffle plate, in conjunction with the increased blowdown capacity, will reduce the amount of sludge that can accumulate in the generator.

Full depth expansion of the tubes in the tubesheet is an improvement over the existing partially expanded aranagement and will minimize both crevice boiling and buildup of impurities in the tube to tubesheet crevice region.

A quatrefoil support plate design will be used in the repaired steam generators.

In contrast, the existing steam generators use drilled hole support plates which have a very limited opening between the tube and tube support plate.

The majority of flow in this drilled plate design is through separate circulation holes.

The tube denting phenomenon, discussed

earlier, has occurred when corrosion products (magnetite) have built up in the tube/ tube support plate intersections (annuli) to the extent that the gap between the tube and support plate closes completely.

The broached or quatrefoil design has no separate circulation holes.

Substantial flow and much flow velocity will take place through the large open spaces in the quatrefoils around, each tube.

This results in a continuous flushing action, tending to wash out this tube/tube support plate area and thus prevent sludge deposits or scales.

The quatrefoil support plate design has led to some tube degradation, in the form of a type of erosion cavitation mechanism, in once-through steam generators.

Although FPL has suggested that this will not be a problem in recirculating designs, we feel that the phenomenon is not understood well enough to assume that recirculating type designs will not see this type of degradation.

Despite this reservation and for the reasons discussed above

3-2 with regard to tube denting, we concur that the quatrefoil support plate design is an improvement over the existing drilled hole design and should be less prone to denting.

The repaired steam generators will use SA-240 Type 405 ferritic stainless steel for both the tube support plates and flow distribution baffle plate.

-The corrosion data provided indicate that, under the test condition, Type 405 stainless steel will be a greatly improved material for tube support plates over the carbon steel presently used.

In the event that denting reactions be initiated, we would have some concern over the propensity of this material for stress corrosion cracking in a chloride environment.

However, Westinghouse appears to have taken the proper precautions in stress relieving it to minimize the likelihood that stress corrosion will occur in the absence of denting.

The Inconel 600 tubing will be thermally treated, which should result in improvement in its resistance to stress corrosion cracking in the reactor coolant and secondary water, particularly in the U-bend regions.

Further, in the eight innermost rows of tubes, the U-bends will be stress relieved after bending.

We find this residual stress relieving process to be satisfactory and an improvement over existing practice.

SUMMARY

Based on the information discussed and the evaluation made above, we conclude that the structural, mechanical, and materials aspects of the FPL proposed steam generator repair program are acceptable and there is reasonable assurance that the health and safety of the public will not be endangered.

We further conclude that the new steam generator design has incorporated features to eliminate the potential for various forms of tube degradation observed to date.

3.2 3.2.1 Effects of Re air Activities Protect>on of Safet Related E ui ment FPL will take measures and establish controls to prevent construction accidents and protect safety-related structures, systems and components from the hazards associated with steam generator transportation and repair activities.

The general precautionary measures that will be taken by FPL include the following:

1.

The reactor vessel will be completely defueled prior to the repair work.

2.

The entire repair process will be preplanned to assure that it can be completed safely and efficiently.

3"3 3.

The repair program will be carried out in accordance with the FPL cor porate Quality Assurance Manual (FPS-NQA-100) and Section Xl of the ASME Code.

Bechtel Corporation has been retained by FPL as the Architect Engineer for the repair program.

4.

The containment boundary will not be disturbed except to open the equipment hatch.

5.

The existing polar crane trolley will be replaced by a higher capacity temporary construction hoist.

The temporary hoist will be inspected and tested prior to its use for construction lifts on the polar crane bridge and the removal of the steam generators.

Defueling of the reactor will begin shortly after shutdown and the normal procedures for defueling will be followed.

The fuel will be stored in the spent fuel storage pool for the duration of the outage.

The temperature of the pool is normally maintained at 95 F or less when the pool contains all of the fuel from the core and the spent fuel elements currently being stored.

We independently estimated the cooling capability of the fuel pool cooling system in its evaluation of the increased storaged capacity of the pool.

With our assumptions, including transferring the fuel 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown, the calculation indicated that the fuel pool temperature would not exceed 139'F.

The spent fuel cooling system consists of two redundant trains designed to seismic requirements.

If a failure were to disable one of these trains, the remaining train could maintain the pool water temperature below 160 F.

We find that these temperatures are acceptable.

In addition, if all of the system cooling of the fuel pool of the unit under repair were to be lost, the fuel pool could still be cooled by the operating unit cooling systems.

The component cooling water (CCW) system of the operating unit has sufficient capacity to supply operational cooling needs, including its spent fuel pool, as well as the cooling needs of the spent fuel pool of the unit under repair, through the existing piping inter-ties between Unit 3 and 4.

Moreover, based on our independent analysis, if the spent fuel pool cooling were to fail following a fuel core off-load, the heat-up rate would be such that boiling of the pool water would take S-l/2 hours.

This is sufficient time in which to make repairs or find an alternate source of make-up water for the spent fuel pool.

Therefore, the present cooling capacity of the spent fuel pool and available makeup sources is adequate for the complete defueling of the reactor as planned for the steam generator repair activities.

In addition, specific potential hazards considered by FPL included the dropping of a steam generator lower assembly, a transporter

accident, the toppling of a crane, the interaction of systems shared by both units and

fires, each of which is discussed below.

3-4 In assessing potential hazards associated with the transportation of the steam generator lower assemblies, FPL considered failures of the trans-porter which consists of a semi-trailer and a haul vehicle.

FPL considered structural failure, overturning, and road failure.

In considering over-turning, the licensee found that it would require the loaded trailer bed to be inclined beyond a

31 angle from the horizontal.

The planned side slopes of the haul route are far less than this 31'ngle.

Further, administrative limits will be placed on the turning radius and speed of the transporter to preclude overturning.

The roadway along the haul route has been evaluated and appropriate sections will be upgraded in order to preclude roadway collapse or damage to the facilities that pass under it, such as electrical duct banks and intake cooling water lines.

FPL has considered the consequences of dropping a steam generator assembly (the heaviest load to be lifted during this repair program) either inside or outside the containment building.

Since there will be no fuel in the containment building while heavy loads are being lifted, there will be no hazard associated with fuel assemblies.

FPL has evaluated the consequences of a postulated drop of the 205 ton steam generator lower assembly on buried facilites along the haul route.

These include intake cooling water piping and electrical duct banks.

Because of the existing cooling water interties between the two reactor units, the cooling systems would be re-aligned as necessary to provide cooling to a possibly damaged cooling system of one of the units.

In the event of damage to the local control cables, alternate starting procedures for the affected pumps are available.

Mith regard to dropping a steam generator assembly outside of the containment building, no other safety related structures (such as the radioactive facility and the fuel storage building) are within the range of the devices used to lift the steam generators from the equipment hatch platform to the transporter.

Based on our review of the FPL consideration above we have concluded that dropping a steam generator lower assembly will present no undue risk to safety-related structures.

FPL considered the toppling of a crane having a 70 foot boom.

The potential consequences of such an accident were considered with respect to the safety-related structures, systems and components of the operating unit.

The diesel-generator building and the auxiliary building were determined to be able to withstand the boom impact without penetration that would result in damage to equipment necessary for the safe shutdown of the operation unit or, in the case of the auxiliary building, the maintaining of the spent fuel pit cooling system.

During the repair the fuel is removed from the affected containment building so that a toppling of the crane on this containment would not present a safety problem.

Damage to the refueling water storage tanks and the primary water storage

tanks, located along the proposed haul route, is precluded since the crane boom will be in the lowered position while traversing these roads.

Based on our review of the FPL considerations, we have concluded that the falling of the crane boom on these safety related structures would not prevent the safe shutdown of an operating unit and would not prevent adequate cooling of the fuel assemblies in the spent fuel pool.

3"5 3.2.2 Other Interactions with the 0 eratin Unit The normal and emergency electrical power distribution systems were reviewed to ensure that construction loads will not jeopardize the supply of electrical power to the operating unit.

The results of that review are discussed below.

3.2.2.

1 Offsite Power S stem The offsite power supply system consist of two start-up transformers to the Turkey Point Units 3 and 4.

Each of the two units has a dedicated start-up transformer which can automatically supply all AC power to both safety non-safety loads of each unit.

Each start-up transformer is capable of supplying the auxiliary loads for its associated nuclear unit and the safety loads for the other nuclear unit.

The temporary loads which are required for the repair of a steam generator will be fed from a temporary 1500 kva transformer.

After the reactor has been brought to cold shutdown the temporary transformer wi 11 be energized by the non-safety

4. 16kv supply system through the start-up transformer to the switchgear of the reactor coolant pump of the unit under repair, the onsite electrical distribution system will be configured the same as during a normal plant refueling shutdown.

On this basis we conclude that the temporary electrical system modification will not degrade the onsite power system in the operating unit.

A fault in this temporary load distribution addition will not cause a loss of power on the reactor coolant pumps in the operating reactor.

3.2.2.2

, Emer enc Power S stem A.

The onsite. emergency diesel-generator system for Turkey Point Units 3 and 4 consist of 2 diesel-generators.

The two diesel generators supply the emergency power to the Turkey Point Units 3 and 4.

The diesel-generators start on either a safety injection signal or on the loss of voltage on a 4160V bus(es) of either unit.

Upon loss of voltage, the following automatic sequence is initiated:

(1)

Diesel-generators are started; (2)

"Preferred supply" breakers of the 4160Y buses are tripped; (3)

Diesel-generator supply breakers close.

In case of a safety dence with the loss sequential starting operating unit.

injection signal on the operating unit in coinci-of power, step 3 above is followed by the of all engineered safeguard equipment for the In case of a safety injection signal in the operating unit, without loss of power, both diesel-generators are started and maintained in an idling mode.

3"6 During the repair of the steam generators, the engineered safety features (ESF) equipment of the unit under repair will be disabled after the reactor has been defueled by the associated feeder breakers being locked open and tagged.

In addition to the lockout of power to the ESF loads connected to the buses of the unit under repair, all those buses that can carry any initiation signal to the shared diesel-generators and which could potentially cause them to become dedicated to the unit under repair (and its loads) will disabled by disconnection.

This step is necessary in order to prevent any possibility that the shared diesel-generators,.and its loads, may become dedicated to the unit under repair Me find these provisions, proposed by FPL to ensure dedication of on-site emergency power to the operating unit, acceptable.

Upon completion of the steam generator replacement work in each unit, the circuitry is to be tested for proper performance prior to the resumption of power operation for that particular nuclear unit.

Mith regard to the power requirements for the spent fuel pool cooling, we have determined that emergency

power, assuming a total loss of off-site power, is not required to be available in less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for any safety functions.

FPL has confirmed that power could and would be restored within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> by manual operator actions to the fuel pool of the unit under going repair.

Me have required that FPL prepare procedures to ensure this latter capability.

B.

The diesel fuel-oil storage system for the two diesel-generators at Turkey Point Units 3 and 4 consists of two day tanks within the diesel generator buildings and one main storage tank outside the building.

Inasmuch as the day tanks have a combined capacity of only 8000 gallons the main storage tank must be operational in order to meet the Technical Specifications for the plant, which require that there be an avail-ability of 40,000 gallons of diesl fuel-oil.

During the construction phase of the steam generator repair program the containment ramp will be removed and replaced by a temporary loading platform.

Inasmuch as the containment ramp of Unit No.

3 is a part of the oil retention dike around the main storage tank, the removal of the ramp eliminates the fire protection feature of'he dike.

In view of this, some remedy is needed in order to restore the main fuel oil storage facility to its fully available condition.

One alternative is to replace the missing portion of the dike with a temporary structure, and the other alternative is to drain the diesel fuel-oil from the main storage tank and place the fuel-oil in a temporary location elsewhere on the site.

If the fuel oil is placed in a temporary location, the supply must be verified to be operational prior to disabling the permanent system.

Either of these alternatives is acceptable in concept.

However, the final choice by FPL must be designed to assure that the Technical Specifications of the plant are satisfied and that the choice meets

3"7 the minimum NRC standards and requirements associated with the operating license.

This will include appropriate application of quality assurance and seismic site requirements to any temporary str'uctures, piping and components; of cleanliness requirements on the fuel-oil; and of other existing functional and operation requirements of this fuel oil supply.

Me require that these details be addressed and adequately demonstrated by FPL prior to initiating the construction changes affecting the fuel-oil retention dike surrounding the main diesel fuel-oil storage tank.

C.

Summary 3.2.3 The spent fuel pool emergency power requirements are acceptable on the basis that FPL submits acceptable procedures to ensure that power can be restored to the spent fuel pool of the unit under repair within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

The diesel oil fuel storage supply is acceptable, however the details of the FPL plan to assure the diesel fuel supply during the repair of Unit 3 must be addressed and adequately demonstrated prior to initiating the construction changes affecting the fuel oil supply.

On the basis of our above review and the satisfactory resolution of the conditions cited we conclude that the provisions by FPL to ensure dedication of onsite emergency power to the operating unit are acceptable.

Fire Protection An evaluation of the fire protection program for the Turkey Point plant Units 3 and 4 containment buildings was included in the NRC Safety Evaluation dated March 21, 1979~~.

This information is supplemented by the FPL report'Steam Generator Repair Report for the Turkey Point Power

Station, Units 3 and 4",~ which addressed the specific fire hazards associated with the steam generator repair outage.

In this regard it should be noted that a fire inside containment cannot cause off-site radioactivity exposures of consequence because the fuel will be removed from the containment of the unit under repair nor can it impair the safe shutdown capability of'he plant.

Nevertheless, the following is a summ-ary of the fire protection measures to be taken during the repair operations.

The use of combustibles in the containment will be minimized to the extent practicable.

Fire retardant scaffolding and materials will be used.

Good housekeeping will assure that wood crates and other combustible trash are removed from the containment in a timely manner.

However, additional amounts of combustible materials will necessarily be introduced into containment including protective clothing, cleaning fluid, charcoal filters and plastic sheeting but the use of these will be mini-mized in those areas-in which cutting and welding is being done.

3-8 The fire protection for the containment consists of fire extinguishers throughout the containment and portable fire etinguisher's will be accessible in the work areas when cutting and welding is performed.

A portable foam system suitable for use inside containment on liquid hydro-carbon fires will be on site and promptly available upon demand throughout the repair.

The existing containment lighting system and emergency lighting are available.

3.3 3.3.1 Even though FPL will not provide a permanently-installed fire water stand-pipe system in each containment before the initiation of the steam gener-ator reapir program a fire hose of sufficient length to reach the most remote-steam generator compartments will be available and dedicated to fight fire inside containment.

A fire watch will be continually present during all welding and cutting operations.

Administrative controls related to fire protection are presently in effect at the plant and are applicable during the steam generator repair outage.

Additional fire protection personnel will be assigned to the replacement activities in the containment.

All administrative site procedures will be reviewed for the control of combustibles and these procedures will identify all known potential fire hazards.

A fire plan for the repair activities will be formulated and coordinated with the station fire plan.

Based on our review of the fire protection measures to be taken to protect safety-related structures, systems and components, we have concluded that there is reasonable assurance that the proposed construction activities can be conducted without significantly increasing the potential for damage to safety-related systems.

Transient and Accident Anal ses Discussion This section discusses the effect the repaired steam generators have on the transient and accident analyses.

As can be seen from Tables 3.3-1 and 3.3-2, FPL has stated that the majority of the relevant design parameters and plant operating parameters will not be changed from those for the present steam generators during steady state.

Therefore systems responses to transient conditions with the repaired steam generators are expected to be essentially the same as for the original steam generators prior to tube plugging.

The impact on the transient and accident analyses is, therefore, not, significant and FPL analyses presented in Final Safety Analysis Report (FSAR) remain valid for the repaired steam generators.

In the following sections we have discussed possible changes in the events previously analyzed in the FSAR.

The following plant conditions were used in those analyses:

3-9 TABLE 3.3-1 STEAM GENERATOR DESIGN DATA PER STEAM GENERATOR)~

Design Pressure, Reactor Coolant/Steam, psig Reactor Coolant Hydrostatic Test Pressure (tube side), psig Hydrostatic Test Pressure, Shell Side, psig Design Temperature, Reactor Coolant/Steam, degrees F

Steam Conditions at 100K load, Outlet Nozzle:

Steam Flow, lb per hr Steam Temperature, degrees Fahrenheit Steam Pressure, psig Feedwater Temperature at 100K load, degrees Fahrenheit Overall Height, fit-in.

Shell OD, upper/lower, in.

Shell Thickness, upper/lower, in.

U-tube OD, in.

Tube Wall Thickness (nominal) in.

Number of Manways/ID, in.

Number of Handholes/ID, in.

Number of U-tubes Tube length (largest U-bend), in.

Total Heat Transfer Surface Area,3ft Reactor Coolant Mater, Volume, ft Reactor Coolant Flow, 1b/)r Secondary Side Volume; ft tSecondary Side Mass No Load, lbs fSecondary Side Mass 100K Power, lbs Center of Gravity (from the support pads),

ft/in.

~0ri inal 2485/1085 3107 1356 650/556 3.2 x 10 516. 0 770 436. 5 63" 1. 6 166/127 3.5/2.63 0.875

0. 050 4/16 2/6 3260 397.5 44,430 945 33.83 x 10 4580 134,000 76,300 25/4 Refurbished N. C."

N.C.

N. C.

N. C.

N. C.

N. C.

N. C.

N. C.

N. C.

N. C.

N. C.

N. C.

N. C.

N. C.

6/6 3214 (~ -1.4X)

N. C.

43,467

(~ -2.,2X) 935 (~ -1.1X)

N. C.

4596 (

40. 3X)

N. C.

80,300 (

+5. 2X)

N. C.

"No change tValues are rounded off

3-10 TABLE 3.3-2 COMPARISON OF PARAMETERS FOR ORIGINAL AND REPAIRED STEAM GENERATORS Primary Pressure Drop Fouling Factor "Nominal Flow Area Equivalent Tube Length Total Heat Transfer Surface Area Heat Transfer Coefficient Nominal Power/SG Nominal Hot Leg Temperature Nominal Cold Leg Temparature Decreased by 0.7 psi Unchanged Decreased by ~1.5X Unchanged Decreased by ~2.2X Increased by ~2.5X Unchanged Unchanged Unchanged

• This decrease in flow area: is due to the reduction in number of steam generator tubes.

Credit has not been taken for the compensating increase in flow area due to the improved. manufacturing tolerance on the tube wall thickness.

3"11 Thermal design flow, gpm/loop S.

G. tube plugging, X

"Power level, Mwt (100K)

T at 100~ power, F

bT at lOOX power, 'F Steady state DNBR N

Fb,H F~ maximum 89,500 0

2200 574.2

55. 9

l. 63

l. 75

2. 55 "The analyses conservatively used 102K power (2244) and T

+4'578.2) avg It should be noted that for this evaluation the FSAR constitutes the reference cycle.

Therefore, if the values of any core physics or plant operating parameters for the reload cycle following the steam generator repair are not bounded by those used in the

FSAR, a reevaluation of the affected event(s) will be required prior to operation.

Any such reanalyses submitted to the staff should be in accordance with Regulatory Guide 1.70, Revision 3.~~

3.3.2 It'hould also be pointed out that the current Emergency Core Cooling System (ECCS) analysis of record for the plant using an approved model is only for the current condition of the original steam generators, i.e.,

with plugged tubes.

If credit for the unplugged configuration of the repaired steam generators is to be taken, a new ECCS analysis using the approved model will be required.

A reload report will be submitted for our review and approval prior to startup of the repaired unit if the fuel loading is different than previously reviewed.

Non-LOCA Accidents and Transients In our evaluation, only the potential effects of the repaired steam generators on the FSAR analyses have been considered.

All other parameters:

are'assumed to have their FSAR values.

As will be seen, most events are not affected by the slight changes which have been made to a few of the relevant parameters.

For some events, such as Rod Cluster Control Assembly (RCCA) withdrawal and RCCA ejection, there will be no effect due to the repair of the steam generators.

The nuclear and thermal time constants of the fuel are much smaller than the fluid mixing and transport time.

These events are termi-nated in less than a loop transport time and, therefore, are unaffected by the steam generators.

For the RCCA drop accident and the malpositioning of part length rods (note that removal of these part length rods has been approved by the NRC

) the neutron flux redistribution is the limiting consideration.

Since this is not dependent on the steam generator performance these analyses are not affected.

3"12 For the loss of reactor flow events, the reactor is rapidly tripped on low frequency, low voltage or low coolant flow.

Changes in coolant temperature due to secondary parameter changes would not be detected in the core during the time frame of interest, for these events.

These analyses are, therefore, also unaffected.

For a chemical and volume control system malfunction, the boron dilution rate depends on the charging pump characteristics and the reactor coolant volume.

The small reduction in reactor coolant volume (~lX) from the FSAR value will not significantly change the time available for operator action.

Therefore, this minor design change will have a negligible affect on the analysis of this event.

The turbine generator design analysis is not affected by the repair of the steam generators since steam and feedwater conditions are unchanged.

The steam generator repair may affect those events for which the transient reactor coolant conditions result from an interaction of the reactor coolant with the secondary system.

These remaining events, which are generally concerned with coolant heatup or cooldown through the secondary

side, are discussed in the following sections.

For the repaired steam generators the increase in the heat transfer coefficient (U) offsets the decrease in heat transfer area (A) so that the resulting heat transfer (UA) remains essentially unchanged.

3.3.2.

1 Excessive Load Increase This event involves a rapid increase in steam generator steam flow which causes a power mismatch between the reactor core power and the steam generator load demand.

This results in a decrease in reactor coolant'emperature and increase in core power.

The FSAR analysis shows that a 10X increase in steam flow from full power can be accommodated without reactor trip.

The repaired steam generators, which have a higher (~5K) full power fluid inventory, could cause the transient to progress slower.

However, the same final steady state condition will be reached.

3.3.2.2 Startu of an Inactive Reactor Coolant Loo This event involves the injection of colder water into the core and a signi-ficant increase in core flow.

This results-in a rapid increase in core power.

The FSAR analysis assumed that the water in the inactive loop was at the saturation temperature of the secondary side.

This is independent of the heat tr ansfer characteristics of the steam generator and will, therefore, be unchanged.

The reduction in reactor coolant volume would cause a negligible reduction in the duration of the cold water slug.

The delay time for the slug to reach the core will remain unchanged.

Therefore, the FSAR analysis of this event would not be significantly affected by the repaired steam generators.

3-13 3.3.2.3 Excessive Heat Removal Due to Feedwater S stem Malfunction This event involves the addition of excessive feedwater to the steam generator or the inadvertant opening of the feedwater bypass valve.

This results in a decrease in reactor coolant temperature and an increase in core power due to moderator feedback.

At full power, the FSAR analysis shows that a new steady state condition is reached without reactor trip.

Since the repaired steam generators will have a higher full power secondary side mass inventory, the cooldown rate would be slower.

However, the same endpoint condition will be reached.

3.3.2.4 Loss of External Electrical Load A loss of external electrical load event such as a turbine trip results in an increase in reactor coolant temperature and pressure and a decrease in core power.

The complete loss of load from 102 percent power analyzed in the FSAR assumed that there was not a direct reactor trip due to the turbine trip.

The increase in full power inventory of the repaired steam generators would provide additional heat capacity and reduce the heatup rate.

Therefore, there are no adverse effects on this event due to the repair of the steam generators.

3.3.2.5 Loss-of Normal Feedwater The loss of normal feedwater results in a loss of capability of the secondary system to remove the heat generated in the core.

Since the repaired steam generators will have a higher full power secondary side mass inventory, additional steam generator heat removal capacity is available; Also, since the dimensions of the steam generators have not changed,. the FSAR conclusions that the tubesheet in the steam generators receiving auxiliary feedwater will remain covered and adequate heat transfer capability will be maintained remain valid.

Therefore, there are no adverse effects on this event due to the repaired steam generators.

3.3.2.6 Loss of All AC Power to the Station Auxiliaries The loss of AC power with turbine trip and reactor trip results-in a reactor coolant flow coastdown to natural circulation flow rates and an increase in secondary pressure.

In the repaired steam generators the tubes will be recessed slightly into the tubesheet

holes, thus reducing pressure drop at the entrance to the tubes which will enhance flow.

Therefore, the FSAR analysis of this event is conservative for the repaired steam generators.

3. 3. 2. 7 Ru ture of a Main Steam Pi e

A steamline break results in a rapid depressurization of the steam generator, a decrease in reactor coolant temperature, and an increase in core reactivity.

The FSAR analysis was performed for end of cycle, hot shutdown conditions.

This event is unaffected by the repair of the steam generators because the no load fluid inventory of the steam generators which was used in the FSAR is still bounding, and the flow area of the main steam line, the reactivity coefficients and the emergency shutdown system are unchanged.

3-14 3.3.3 Loss-of-Coolant Accident LOCA) 3.3.4 The minor design and operational differences of the repaired steam generator, such as number of tubes, full power fluid inventory, and pressure drop across the steam generator, are not expected to significantly affect the LOCA analysis.

The reduction in flow area and reactor coolant volume due to the-lesser number of tubes is approximately equivalent to 1.4X of the tubes in the original steam generator being plugged.

The FSAR ECCS analysis is based on a model which the staff no longer finds acceptable.

Therefore, the analysis cannot be used to satisfy the require-ments of 10 CFR'0.46.

As mentioned

above, the ECCS analysis of record, based on the currently approved

model, has been performed assuming a

significant number of steam generator tubes plugged.

We consider the ECCS analysis of record to be conservative for plant operation with the repaired steam generators.

If credit for the unplugged configuration of the steam generators is to be taken, a new LOCA analysis performed with the currently approved model must be submitted.

The repaired steam generators do not have a significant effect on the small break LOCA.

Therefore, the current small break LOCA analyses are acceptable for the plant with the repaired steam generators.

Steam Generator Tube Ru ture 3.3.5 The improved manufacturing tolerance on the tube wall thickness will result in a slight increase in the tube inner diameter.

This increase in diameter (0.005 inch) will have a negligible affect on the tube rupture analysis.

Therefore, the consequences of this event, as reported in the FSAR, will be unchanged by the steam generator repair.

~Summa r The changes in design and plant operational parameters listed in Tables 3.3-1 and 3.3-2 have been evaluated to determine their effect on the safety analyses'.

We have concluded that the repaired steam generators will not have any significant adverse effect on the transient and accident analyses and therefore, that the analyses and conclusions presented in the FSAR (except for LOCA) remain valid for the same core physics and plant operating parameters.

For the LOCA, new analyses will be submitted as discussed in Section 3.3. l.

3.4 3.4.1 Radiolo ical Conse uences of Postulated Accidents Ace>dents Ourin 0 eration with Re a>red Steam Generators The repaired steam generators will not significantly affect the dose con-sequences of accidents involving the secondary system.

The accidents involving significant dose consequences are the main steam line failure, steam generator tube failure and control rod ejection.

The only design change that affects the accident dose consequences is a 5X increase in the

3-15 volume of the secondary side of the steam generator.

The reactor coolant system parameters which affect these accidents will not be changed signifi-cantly by the repaired steam generators.

These parameters include reactor coolant leakage to the secondary system and the reactor cooldown period.

The contribution to offsite doses from the secondary system is minor in all three accidents because of low activity levels in the seconary system.

The major dose contribution is from reactor coolant leakage into the secondary system during the accidents.

In both the steam generator tube failure and control rod ejection accidents, the increased volume of the secondary system provides for more dilution of the-activity which leaks from the reactor coolant site.

Because the reactor coolant system parameters have not changed, the total reactor coolant side release time and volume will not change.

Therefore, the increased secondary volume should result in a negligible change in doses.

The reactor coolant system parameters which affect the main steam line failure accident also remain unchanged.

Assuming the same concentration of radionuclides (pre-existing inleakage of reactor coolant),

the increased mass of the secondary side will result in a slight increase in offsite doses.

The contribution to the doses from additional reactor coolant inleakage during the accident itself would be unchanged.

Because the secondary volume increases by 5 percent and most of the dose is a result of "fresh" reactor coolant inleakage, the total offsite dose will increase by much less than 5 percent.

This slight increase in total offsite dose will not reult in estimated consequences in excess of the 10 CFR Part 100 guidelines, and the conclusions concerning these accidents reached in the Harch 15, 1972 Safety Evaluation for the Turkey Point Plant~~ are not changed due to the repair of the steam generators.

3.4.2 Accidents Ourin the Re air Effort FPL has analyzed the potential consequences of postulated accidents associated with the repair effort.

FPL has analyzed the potential for steam generator crane rigging accidents which may affect the refueling water storage tank and primary water storage tank and concluded that rigging operations will be conducted in areas sufficiently removed from these tanks to preclude damage to these structures.

FPL has also evaluated the potential for a steam generator being dislodged from the rigging and striking the radwaste or fuel handling building.

FPL has concluded that both buildings are capable of withstanding all postu-lated impacts with no breach of integrity.

Me have evaluated the FPL report~

and concur with the above conclusions.

Therefore, we conclude that there will be no radioactive release to the environment from these construction related accidents.

,FPL has analyzed the potential consequences of rupturing the steam generator boundary due to mechanical shock and concluded that even if the primary side boundary is breached, the tenacious nature of the corrosive film would result in insignificant releases to the environment.

3-16 We have independently analyzed the potential consequences of a steam generator drop.

We have assumed that dropping of a contaminated steam generator could rupture the reactor coolant side boundary, thus exposing the contaminated reactor coolant side surfaces.

It is expected that most of the activity on the reactor coolant side is tightly bound to the piping surfaces.

This is evident by the fact that the activity was not removed by the high velocity reactor system flowrates during operation.

Radio-activity which may become loosened due to the drop will mostly be deposited on the large surface areas inside the steam generator lower plenum because there will be little air movement between the steam generator internal air spaces and the outside atmosphere.

Consequently, we have conservatively assumed that 0. 1 percent of the activity in the steam generator becomes airborne and is released to the atmosphere.

The resultant dose to the critical organ of an individual at the site boundary is 0.02 rem to the lung.

The assumptions used in the calculation and the results are given in Table 3.4-1.

3.5 S ecial License Conditions During the repair program the following temporary license conditions will be imposed:

(1)

All fuel shall be removed from the reactor pressure vessel and stored in the spent fuel pool.

(2)

The health physics program and procedures which have been established for the steam generator repair program shall be implemented.

(3)

Progress reports shall be provided at 60 day intervals from the start of the repair program and due 30 days after close of the interval with a final report provided within 60 days after completion of the repair.

These reports will include:

(i)

A summary of the occupation exposure expended to date using the format and detailed of Table 3.3-2 of the "Steam Generator Repair Report" as supplemented.

(ii) An evaluation of the effectiveness of dose reduction techniques as specified in Section 3.3.5 of the "Steam Generator Repair Report" as supplemented in reducing occupational exposures.

3"17 TABLE 3.4-1 ASSUMPTIONS USED IN CALCULATING RADIOLOGICAL CONSE UENCES OF STEAM GENERATOR DROP Activity in Steam Generator (Ci)~

Fraction of Activity Becoming Airborne Site Boundary X/g (S/m )

Lung Inhalation Dose Conversion Factor "" (

. )

pCi m

Breathing Rate (')

S 1400 0.001 5.5 x 10 7.46 x 10-4 3.47 x 10-4 Site Boundary Radiological Consequences of Postulated Steam Generator Drop 0.02 rem All activity is assumed to be Co-60.

From-Regulatory Guide

1. 109.

3-18 (iii) An estimate of radioactivity released in both liquid and gaseous effluents.

3 ~ 6 (iv) An estimate of the solid radioactive waste generated during the repair effort including volume and radioactive content.

(4)

Procedures shall be prepared to assure that power can be restored by manual operator actions to the fuel pool of the unit undergoing repair within eight hours.

(5)

The remedy chosen by FPL to provide the availability of the diesel fuel supply while the oil retention dike is removed from the main diesel safety tank shall'e addressed and adequately demonstrated by FPL prior to initiating the construction changes affecting the dike.

(6)

Sixty days prior to fuel loading, the program for preoperational testing and startup shall be submitted for NRC review.

~Securit FPL has an approved Hodified Security Plan~~ which will be implemented during the repair program to assure that the security program in effect at the Turkey Point Plant is not degraded as a result of steam generator repair program activities.

Me have reviewed the FPL program in light of these measures and have concluded that the program will not be degraded.

=A

CONCLUSION j

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 regula-tions 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.

Oate:

MAY 1 4 1979

5-1

5.0 REFERENCES

2.

"Steam Generator Repair Report Turkey Point Units 3 and 4" Florida Power and Light Company, September 20,

1977, as supplemented on December 20,

1977, March 7, April 25, June 20, August 4, and December 15,

1978, and January 26, 1979.

Steam Generator Repair

Program, Surry Power Station Unit Nos.

1 and 2, Virginia Electric and Power Co., August 17, 1977 and revisions dated December 2,

1977; April 21, June 2, June 13, June 30, September 1,

October 25 and November 10, 1978.

3.

6.

7.

8.

9.

10.

Safety Evaluation Report by the Office of Nuclear Reactor Regulation, Licensee Nos.

DPR-32 and DPR-37, Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Docket Nos.

50-280 and 50-281, U.S.

Nuclear Regulation Commission, December 15, 1978.

Environmental Impact Appraisal by the Office of Nuclear Reactor Regula-tion, License,Nos.

DPR-32 and DPR-37, Virgnia Electric and Power Co.,

Surry Power Stations, Units 1 and 2, Docket Nos.

50-280, 50-281, UPS.

Nuclear Regulatory Commission, January 20, 1979.

Hoenes, G. R., Waite, D.

A. and McCormack, "Radiological Assessment of Steam Generator Removal and 'Replacement,"

NUREG/CR-0199, PNL-2656 Battelle Pacific Northwest Laboratories for U.S. Nuclear Regulatory Commission, September 1978.

Final Environmental Statement related to the Operation of the Turkey Point Plant, U.S. Atomic Energy Commission, July 1972.

Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE CODE), Office of Standards Development, U.S. Nuclear Regulatory Commission, NUREG-0017 April 1976.

"FPL Quality Assurance Topical Report" FPLTQAR 1-76A, Revision 2, Florida Power and Light Company, September 8,

1977.

"Bechtel Quality Assurance Program for Nuclear Plants," Topical Report BQ-TOP-l, Revision 2A, August 15, 1977.

"Quality Assurance Plan - Westinghouse Water Reactor Division," Topical Report SCAP-8370, Revision 8A," Approved by letter NRC (Heltemes) to Westinghouse (Eicheldinger),

September 10, 1977.

Letter from NRC (Schwencer) to FPL (Uhrig) transmitting amendment Nos.45 and 37 to License Nos.

DPR-31 and DPR-41 for the Turkey Point Plant Unit Nos.

3 and 4 with attached NRC Safety Evaluation for the FPL Fire Protection Program March 21, 1979.

5-2 12.

Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (LMR Edition), Office of Standards Development, U.S.

Nuclear Regulatory Commission, Regulatory Guide 1.70, Revision 3, November 1978.

13.

Safety Evaluation by the Division of Reactor Licensing, U.S. Atomic Energy Commission, March 15, 1972.

14.

Letter from NRC (Schwencer) to Florida Power and Light Co. (Uhrig) transmitting Amendment Nos.

44 and 36 to license nos.

DPR-31 and DPR-41, approving the Modified Amended Security Plan, dated February 27,, 1979.

15.

Letter from NRC (Schwencer) to Florida Power and Light Company (Uhrig) transmitting Amendment Nos.

37 and 30 to License Nos.

DPR-31 and DPR-41, approving the removal of the part length rods, dated September 21, 1978.

16.

"Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," U.S. Nuclear Regulatory Commission, Regulatory Guide l. 109, October 1977.

17.

"Radioactive Material Released from Nuclear Power Plants (1976),"

T.R. Decker, U.S.

Nuclear Regulatory Commission, NUREG-0367, March 1978.

18.

"Radioactive Material Released from Nuclear Power Plants (1977),"

T.R. Decker, U.S. Nuclear Regulatory Commission, NUREG-0521, January 1979.