ML20023C824

Safety Evaluation Supporting Amend 21 to License NPF-9

ML20023C824
Person / Time
Site:	McGuire
Issue date:	05/05/1983
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20023C819	List: ML20023C818 ML20023C824 ML20023C829
References
NUDOCS 8305180066
Download: ML20023C824 (13)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDHENT NO. 21 TO FACILITY OPERATING LICENSE NPF-9

.D.U.KE POWER COMPANY I.

INTRODUCTION By letter dated February 3,1983, Duke Power Company submitted a report entitled "ItGuire Nuclear Station - Unit 1 Steam Generator Monitoring Programs," which out-lined the specific actions and surveillance programs relative to the McGuire -

Unit 1 Podel D2 steam generator modification. The proposed surveillance and moni-toring program is based on the reconmendations made by the Design Review Panel (DRP) in their January 1983 report "D2/03 Steam Generator Design Modification."

The staf f evaluation (HUREG-0966 March 1983) of the DRP report concluded that the modification of the D2/03 steam generators is acceptable and that the modified steam generators can be operated at 1007. of their design capacity. The licensee also proposed a programnatic license condition to verify acceptability of the modi-fication.

The DRP identified three specific items to be addressed by each of the utility owners installing the proposed preheater modifications.

These items are as fol-lows:

1.

Provisions should be made for ' initial monitoring of inlet pressure oscillations:

2.

Plant-specific provisions for assuring feedwater flow and/or feed-water temperature restrictions are met should be described, where applicable; 3.

Inservice inspection, eddy current testing and tube vibration moni-toring programs and schedules should be described, where applicable.

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The means by which each of the above items will be' implemented on McGuire Unit 1 and the schedule for programs in item 3 are described in the licensee's submittal of February 3,1983, with additional infomation provided in submittals dated March 1, 1983, March 14, 1983, and April 28,1983(revised).

II.

DISCUSSION AND EVALUATION OF PLANT SPECIFIC ITEMS A.

Inlet Pressure Monitoring, In Section 5.2.13 of its report, the DRP recommended that the pressure oscillations in the feedline be initially monitored throughout the design i

operating flow range. To accomplish this, the licensee has proposed a 8305180066 830505 PDR ADOCK 05000369 P

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pressure monitoring program to record data during the power escalation 3

period following installation of the preheater modification.

3 The intent of this monitoring is to monitor all pressure variations which could affect the fetigue usage factors of the bolts and welds.

This is to by accomplished be using a pressure' transducer installed

.in the feedline elbows. Measurements will be made;over the design operating flow range, i.e., from 17% power, where main feed flow is initiated, up to 1005 power. Power escalation will be made in 3%

increments. Measurements will be made during the period that power is increasing as well as at each 31 increment.

The analysis by Westinghouse for steady state pressure fluctuation resulted in the development of curves of allowable peak-to-peak pres-.

sure oscillations versus frequency. These were developed for critical.

modification components most subject to this loading and are based on limiting the oscillating pressure stresses at any frequency to the endurance limit for the material.

Acceptance criteria for this test will be established to verify that the plant measurements fall within the bounding values used by Westinghouse in the analysis of the manifold.

i Based on our review of the proposed program for inlet pressure monitor-ing we find that the licensee has met the requirements of DRP item 1.

We therefore find the program acceptable.

l B.

Feedwater System Changes The McGuire steam generators are provided with separate inlet connections for main feedwater and auxiliary feedwater piping. The auxiliary feed-water system (AFWS) is used to provide makeup to the steam generators during plant startup until steam generator flow requirements approach the AFWS design capability. At this stage of plant warmup, the main feedwater system is actuated. When main feedwater is introduced into the lower main feedwater inlet nozzles, the colder water that has stagnated down-stream of the main feedwater isolation valve is injected into the steam 1-t generator. Westinghouse has calculated that the ensuing themal transient '

will result in an overstressed condition on several of the proposed inlet L

distribution manifold bolts. The problem relates to a combination of low.

feedwater line purge flow and cold feedwater in the feedline between the isolation valve and the steam generator. The DRP recoreended that each i

utility provide some plant specific method to alleviate this situation.

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A bypass line and locked-open manual valve will be provided to bypass the main feedwater check valve at the steam generator to allow backflow into the main feedwater line. Backflow flushing of the main feedwater lines will be performed during plant wamup when the AFWS is being used for. steam generator makeup prior to actuation.of the main feedwater system. Hot effluent from the steam generator will. bypass the check E

valve and preheat the main feedwater inlet piping up to the main feed-water isolation valve outside containment. Backflow will continue l

until the main feedwater piping is adequately preheated as detemined by three themocouples provided inboard of the containment isolation -

valve. -Two new sections of feedwater piping will be added to pemit the bypass backflow to be directed to the condenser hotwell during plant wamup. These new sections of piping will be intarconnected with the existing main feedwster flow tempering line. The existing flow tempering line is used to provide a small main feedwater flow through the AFWS inlet to the. steam generators. One of the new recirculation i.

piping sections will be located in the doghouse (steam and feedwater.

valve compartment) adjacent to the containment and the other new recir-culation piping section and valves will be located in the turbine building. The new bypass line and manual bypass valve will be located l

in the contai ment. An orifice in the new pipeline located in the doghouse will limit backflow through the bypass line to 40 gpm.

1 The piping and valves in the turbine building are nonseismic Category I and have no safety-related function and, therefore, are not protected from

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natural phenomena, including tornado missiles. The bypass valve in the reactor building and the piping between the feedwater piping and flow tempering piping (FW-FT piping) in the doghouse are seismic Category I, l

Quality Group B.

The safety-related portion of the system is located in i

seisnic Category I, flood, and tornado protected structures. Thus, the i

requirements of General Design Criterion 2, " Design Basis for Protection Against Natural Phenomena," and the guidelines of Regulatory Guide 1.29,

" Seismic Classification," Positions C.1 and C.2 are satisfied.. The

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l essential equipment is separated from the effects of internally generated missiles. The applicant indicated that the new components could not

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credibly produce missiles, similar to the original' components and pre-viously reviewed in Sections 3.5.1.1 and 3.5.1.2 of the McGuire FSAR.

j The utility has provided the results of a high energy pipe break analysis j

using the guidelines of the Standard Review Plan Sections 3.6.1 and l

3.6.2 for the new piping. This analysis included additional pipe break locations and the effects of pipe whip, jet impingement, flooding environ-mental effects, and the potential loss of any safety-related equipment in the area. Any safety equipment which is required to operate after the high energy pipe break is protected by shielding from jet impingement and other effects of discharging fluids such as splashing or dripping. There t

i is no moderate energy piping being added by this modification. Thus, the L

requirements of General Design Criterion 4. " Environmental and Missile l

Design Bases," are satisfied.

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This modification is for each unit and there is, therefore, no sharing between units. Thus, the requirements of General Design Criterion S.

" Sharing of Structures, Systems, and Components," are not applicable.

The _feedwater system is not required to transfer heat under accident con-ditions and..therefore, General Design Criteria 45, " Inspection of Cooling Water Systems," and 46, " Testing of Cooling Water Systems," are not applicable. The open bypass line could result in a 40 gpm leakage path around the steam generator check valve. However, the licensee has verified that redundant isolation valves are provided downstream of the bypass line in the return line to the condenser. The redundant' isola-tion valves receive signals to close from corresponding control trains to isolate the return line on automatic start of the AFWS. Therefore, the addition of the bypass line has no adverse effect on minimum AFWS flow i

requirements for any accident previously analyzed. Thus, we conclude that the requirements of General Design Criterion 44, " Cooling Water,"

are satisfied with respect to this feedwater modification not affecting the performance of the AFW system.

Based on the above, we conclude that the modification to the f.eedwater system meets the requirements of General Design Criteria 2, 4, and 44 with respect to its protection against natural phenomena, missile and environmental effects, and in r.ot affecting the performance of the AFW system in mitigating the consequences of an accident, and meets the

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guidelines of Regulatory Guide 1.29, Positions C.1 and C,2, with respect to its seismic classification and is, therefore, acceptable. The modi-

-fication to the feedwater system meets the applicable acceptance criteria of SRP Section 10.4.7.

The licensee, in proposing reverse flushing of the feedline which would eliminate the themal transient causing a high usage factor 'on certain l

' bolts and welds in the modification manifold assembly, has met the requirements of DRP item 2.

The feedwater system piping changes being made by the licensee are therefore acceptable.

C.

Monitoring Program In the DRP evaluation report the DRP recommended that each utility develop inspection, testing and monitoring programs specific to their plant ( s). These programs are designed to verify the hydraulic perfor-mance of the modification and give early indication of any structural problems with the manifold. The DRP's recommended surveillance program included visual inspection of the manifold assembly and baseline ECT of i

the affected first five rows of tubes in the-preheater sections after manifold installation and visual and ECT after a 6 month full. power operational period. Tube vibration monitoring of installed accelero-meters during power escalation was also recomended. The licensee has supplemented the recommended DRP surveillance as described in'the fol-lowing paragraphs.

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1) Visual Inspection I

The visual' inspections proposed by the licensee follow the DRP's

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recommendations..The visual inspections are intended to provide an early indication of any unexpected loss of structural.integ-ri ty. Therefore, a visual inspection of the accessible areas I

of the modified components will be performed.

Inspection access will be through the radiography port in the feedwater piping upstream of the steam generator nozzle. The inspection will be ~

performed using a fiber optics borescope and will be recorded by videotape or still photographs for future reference. Specific i

items to be inspected include bolts and welds for erosion, fretting wear, corrosion anc cracking. The results of the subsequent inspection will be compared with the as-built condi-

- tion of the manifold.. Any questionable or unusual visual'indi-cations will be evaluated to detemine the need for corrective l

action.

If corrective action is required, a report detailing the problem and the corrective action will be submitted to the NRC staff prior to subsequent power operation.

s The visual inspection described above will be performed following I

i reassembly of the feedwater piping after modification installation and again at the first refueling shutdown. The subsequent schedule and the extent of inspection are described in the licensee's April 28, 1983 (revised) submittal.

l The proposed manifold visual examination should be performed in accordance with Boiler and Pressure Vessel Code Section XI IWA-2211 L

Visual Examinations YT-1.

2) Tube t ~tfation Monitoring I

The DRP endorsed the Westinghouse recommendation for tube vibration monitoring for the first plants modified.. Accordingly, four tubes i

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in the 'A steam generator in McGuire - Unit I will be instrumented with two accelerometers each to provide an early indication of manifold performance prior to ed# current testing. Due to the uncertainty in the relationship between tube vibration and wear, no short term acceptance criteria have been: established. However, the results of these measurements are expected to be useful in assessing the long term potential for the manifold to reduce the wear rate to an acceptably small value. Axial location of each accelerometer is given in. Table 1.. Acceleration readings will be recorded for off-line analysis by Westinghouse and the licensee. Limited on-line analysis will be performed to verify the validity of the recorded data.

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TABLE 1 Axial Location of Tube Mounted Accelerometers

, Row Column

• _ Eval uat_1on 49 31 N Halfway between plates 5 and 6 49 31 N Halfway between pictes 6 and 7 49 40 W Halfway between plates 5 and 6 49 40 W At Plate 3 49 60 W Halfway between plates 5 and 6 49 60 W At Plate 7 49 71 N At Plate 3 49 71 N Halfway between plates 5 and 6

• W - window Tube N - Ncn-window Tube Column numbers are those used by Westinghouse. Licensee numbers columns as a mirror image during ECT.

i Data will be recorded during power escalation following installation of the modification. As a minimum, data will be recorded at the following i

power levels during steady state conditions; 40%, 50%, 60%, 70%, 75%,

l 80%, 85%, 90%, 95%, and 100%. Appropriate plant data will be recorded concurrent with the accelerometer data for correlation purposes.

For McGuire Unit 1, the licensee will, in addition to measure.aents taken after startup, take and record data from each of the accelerometers at 100% power during the subsequent operating period approximately halfway between startup and the end of the operational period. These data will be compared with the initial data to verify no significant change in tube behavior.

The design modification is intended to reduce the tube vibration response to acceptable levels, i.e., levels corresponding to 40% power with the original design. The proposed vibration monitoring program shoeild be able to verify that the modification achieves this objective in the tubes 49-31, 49-40, 49-60 and 49-71 to be instrumented with accelero-meters.

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The NRC~ staff and its consultants have reviewed this proposal and concur

with it. In particular, the selected tubes include two window tubes (49-40.and 49-60), two non-window tubes (49-31 and 49-71), a tube on the periphery of the bundle (49-31) which is exposed to " skinning" flow, and i

a central tube (49-60).

It should be noted that tubes 49-40 and 49-71

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were previously instrumented and data were obtained under. operating-conditions with the original design.

Because relatively large tube-to-tube support plate (TSP) hole diametral clearances still remain, the potential for a tube to float within a TSP i~

. exists. ' Further, calculations based on the assumption of uniform flow show that at high power levels fluidelastic instability is possible if a tube can vibrate in a TSP-inactive mode. Therefore, a primary purpose of the accelerometer measurements will be to first detemine if any of the tubes are vibrating in a TSP-inactive mode, and secondly, to detemine if a threshold power level exists above which large amplitudes indicative of l

an instability occur.

The accelerometer data will be recorded on magnetic tape for subsequent data analysis. The data analysis should include frequency spectra, in the fom of power spectral density (PSD) plots, and root mean square (RMS) values which are readily obtained by integration under the PSD curve. Accelerometer signals should be double integrated to obtain E

displacement data. and the PSDs and RMS values should be obtained for l

both acceleration and displacement.

L Dominant frequency peaks should be identified from the PSD curves and com -

pared with results from vibration analyses of the tubes for different assumed support arrangements to detemine if the tube is vibrating in a support-inactive mode. The RMS displacement should be plotted as a func-tion of power level.. An abrupt increase in displacement response.at a -

given power level, coupled with a simultaneous sharpening of the frequency response spectra, is indicative of. a fluidelastic instability. Additionally, the variation of displacement response with power level (flow velocity) can typically be approximated with a power function relationship. -If the exponent on the power levol is on the order of 2-3, one can reasonably assume that the response is caused by turbulent buffeting..On the other hand an exponent of four or more may indicate a fluidelastic instability.

. An assessment of long term potential of the manifold to reduce tube wear rate will be made after the first refueling outage and acceptance criteria for tube vibration will be established and submitted for staff review.

Based on our review of the proposed vibration monitoring program, we find that it has met the requirements relative to vibration monitoring in DRP item 3.

We therefore find the program acceptable.

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3) Eddy Current Testing The primary method for assessing the effectiveness of the steam generator modification in reducing the rate of tube wear will be ed@ current testing (ECT). The same ECT methods will be used for testing after modification installation that were used in previous ECT examinations to' allow comparison of results. The first five rows (45 to 49) will be examined. Al though not specifically required, the peripheral tubes (tubes adjacent to the wrapper) will also be exasined using the same techniques.

3e above inspection will be performed after completion of the modifica-tion on each steam generator. This inspection will serve as the baseline inspection for the modified steam generator. A second ECT examination w',11 be performed after the proposed period of operation. This second examin-ation will include the same tubes examined during the initial.examinaticn.

Subsequent ECT examinations will be performed as required by the McGuire Technical Specifications (which used Regulatory Guide 1.83 for determintag inspection frequency) and as outlined in the licensee's April 28,1983

-(revised) submittal.

4) Loose Parts Monitoring The DRP's recommended surveillance program did not include the use of loose parts monitoring as one means of assuring the continued structural integrity of the installed manifold.. McGuire - Unit i has an installed loose parts monitoring system (LPHS). This system includes a sensor on the lower head of each steam generator. This system, although intended for detecting loose parts in the primary system, has high enough sensitivity to detect a loose l

manifold. Although extremely unlikely, if a signal is detected on the LPHS which indicates that one of the manifolds is loose, the unit will'be shut-down, NRC will be notified and appropriate corrective action taken.

McGuire Technical Specifications require that daily channel checks, monthly operational tests and 18 month calibrations be performed on the LPMS.-

Further, the technical specifications require that the LPMS be operabic.

A report must be suomitted to NRC if any channel is inoperable for more than thirty days.

5) Pluaced Tubes The DRP's evaluation report did not address the presence of plugged tubes t

in a modified steam generator. Operation of McGuire Unit 1 steam generator in the unmodified condition resulted in the plugging of six tubes in November 1982 (one tube which did not have significant wear was plugged in July 1982 due to a misinterpretation of the eddy current signal). These tubes cannot be monitored by eddy current technique directly.

Integrity of these six tubes will be inferred from eddy current information on active l

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If ECT measurements show'that some wear has occurred over several inspection intervals, a wear rate for. the plugged tubes will be estimated.'

If a previously plugged tube is thus evaluated to have reached a defect of 805 through wall, a detailed structural evaluation will be performed to demonstrate its integrity prior to returning to service.

1 The Duke Power Co.'s proposed program of inspection, testing and monitoring will provide sufficient performance verification of the modified steam generators.. The licensee has supplemented the DRP's recommended surveil-lance. program with loose parts monitoring for on line detection of sounds,-

due to foreign objects, loose. parts,' or a loose manifold. emanating from 1

the area of the. installed modification.' The McGuire Unit 1 ECT program.

will also include examination of peripheral tubes in addition to the recom-i mended first five rows of tubes in the preheater section. Tube vibration j

monitoring of the McGuire Unit 1 will include taking and recording.

of accelerometer measurements at half way through the proposed period of 4

power' operation in addition to the measurements recommended to be taken during power ascension to 1001 power.

l As a general reconmendation. ary sensors (such as accelerometers and. loose parts transducer) that were utilized in monitoring tests of the original design steam generator should be left in the same locations for the monitoring / tests of the modified design. Results from the original design can then be used as a baseline against which results with the modified design can be compared.

III.. DISCUSSION AND EVALUATION OF RADIOLOGICAL CONSIDERATIONS l-A.

ALARA Guidelines l

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The staff has evaluated the DRP's radiological assessment of the radiation protection measures established by Westinghouse for the Westinghouse Pre-heat Steam Generator D2/D3 Design Modification, including.those measures intended to ensure that doses will be maintained as low as is reasonably achievable (ALARA). Our assessment is based on the DRP utilization of the criteria outlined in Regulatory Guide 8.8, "Information Relevant to Ensuring that Occupational Exposures at Nuclear Power Stations Will Be As Low As.Is Reasonably Achievable" in the DRP's radiological assessment of the design modification, and its assessments, primarily those provided ~1n Section 4.4,

" Radiological Considerations," Section 5.5, " Radiological Consideration /ALARA,"'

and Section 6.0, " Summary" of the January 1983 DRP Evaluatirn Report. Infor-motion.provided in other sections was also considered whers it contributed to our assessment of the ALARA features of design, planning, installation, maintenance, and inspection. We have additionally evaluated information specific to the McGuire radiation protection /ALARA program which has pre.

viously Aeen submitted in the McGuire Final Safety Analysis Report (FSAR).

This has been evaluated and found acceptable by the staff in our McGuire Safety Evaluation Report (SER).

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D Westinghouse's proposed D2/03 Design Modification Program provides for radt-ation protection /ALxttA measures throughout the design and preparation stage, the performance of the modification, and during post-nodification recovery and operations. The McGuire Unit i radiation protection /ALARA progran has those features essential for compatibility with the Westinghouse design modification program, and contains radiation protection /ALARA elements designed to ensure adequate radiological protection for workers and promote ALARA doses on tasks associated with the modification. These proposed measures are consistent with 10 CFR 20.1(c) and Regulatory Guide 8.8, and are, therefore, acceptable to the !!RC staff for the planned modification of McGuire Unit 1.

B.

Evaluation Based upon our evaluation we find the proposed nessures consistent with 10 CFR 20.1(c) and Regulatory Guide 8.8 and, therefore, acceptable. The licensee performed a radiological assessnent of the proposed modification for the McGuire Unit 1 prior to task initiation to determine the applic-ability of proposed worker radiological protective measures and ALARA con-siderations, and to determine how best to integrate this program with their own facility radiation protection program. A similar assessment will be perfonned for Unit 2.

Where significant differences in any of the radio-logical parameters considered by Westinghouse exist (e.g., equipment, dose rates, radiation sources, doses, training), these will be evaluated and compensating radiation protection /ALARA actions described. During and upon completion of the modification, the ifcensee will perfona a summary radio-logical assessnent of the task, as is recommended in C.3.c of Regulatory Guide 8.8, to enable the staff to evaluate the radiological results of the modification and determine if additional or different radiological controls need to be considered. This will include the following:

(1) The collective occupational dose estimate shall be updated weekly.

If the updated estimate exceeds the person-rem estimate by more than 10%, the licensee shall provide a revised estimate, including the reasons for such changes, to the NRC within 15 days of deter-mination.

(2) A final report shall be provided to the flRC uithin 60 days after completion of the repair. This report will include:

(a) a summary of the occupational dose received by ma.jor task, and (b) a comparison of estimated doses with the doses actually received.

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~IV.

t.ONCLUSIONS We find that McGuire Unit I with modifications to the preheater sections of the steam generators may be operated at full power for a period of approximately 200 days without undue risk to the public health and safety.

Upon shutdown, in addition to the proposed visual inspections and ed@

current tests of preheater tubes, technical specification ed@ current tests of 3% of the steam generator tubes should be conducted.

The proposed manifold visual exaninations should be performed in accord--

ance with Boiler and Pressure Vessel Code Section XI IWA-2211 Visual.

Examinations VT-1.

In addition, all of the visual examinations of the manifold and ECT of the preheater tubes after installation shall be perfomed in accordance with the extent and schedule of examination as specified in the licensee's April 28,1983 (revised) submittal.

In addition we conclude the following:

i 1.

Inlet Pressure Monitoring Duke Power plans to monitor the pressure at the feedline inlet nozzle during the power escalation period following the instal-lation of the inlet modification. The licensee will be monitored pressure throughout the design operating flow range.

In addition, the licensee will verify that the acceptance criteria established by Westinghouse from test data are applicable to the McGuire steam generators and represent bounding and conservative values.

2.

Tube Vibration Monitoring With respect to the placement of accelerometers and data gathering techniques, due to uncertainty in the relationship between tube-l vibration and wear rate, no short term acceptance criteria have-l been established. An assessment of long term potential of the I

manifold to reduce tube wear rate will.be made after the first refueling outage and acceptance criteria for tube vibration will be established and submitted for staff review.

- 3 3.

ALARA a.

Perfonn dose and ALARA pre-modification assessments for McGuire specific actions, 'and :

b.

Provide a post-task sumnary radiological assessment as outlined herein.

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l We also conclude that the license condition proposed by the licensee be included in the McGuire Unit 1 Facility Operating License:

"The licensee shall conduct the inspection, testing and moni-toring program as described in the attachment to Hal B. Tucker's letters of February 3,1983, and April 28,1983(revised). The licensee shall not make any major modifications to this program unless prior NRC approval is received.

" Major modifications are defined as:

a.

Elimination of any identified testing, inspection or moni-

toring, b.

Changes in the frequency of performing the identified test-ing, inspection or monitoring, and c.

Reduction in the scope of any of the identified testing, inspection or monitoring."

We have concluded, based on the considerations discussed above, that: (1)because j

the amendment does not involve a significant increase in the probability or con-sequences of accidents previously considered, does not create the possibility of an accident of a type different from any evaluated previously, and does not _ involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety l

of the public will not be endangered by operation in the proposed manner, and i

(3) such activities will be conducted in compliance with the Commission's regula-tions and the issuance of this amendment will not be inimical to the cosanon defense and security or to the health and safety of the public.

t V.

ENVIRONMENTAL CONSIDERATION We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any signif-icant environmental impact. Having made this detemination, we have further con-i cluded that the amendment involves an action which is insignificant from the stand -

point of environmental impact and, pursuant to 10 CFR 551.5(d)(4).- that an environ-mental impact statement-or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.

Principal Contributors:

L. Frank, Materials Engineering Branch, DE J. Rajan, Mechanical Engineering Branch, DE j

J. Ridgely, Auxiliary Systems Branch, DSI R. Serbu, Radiological Assessment Branch, DSI l

R. Birkel, Licensing Branch No. 4 DL L

Date: May 5, 1983 Npo i

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- 14 of the publ S will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compilance with the Comission's regula-tions and the issuance of this amendment will not be inimical to the common-defense and security or to the health and safety of the public.

V.

ENVIRONENTAL CONSIDERATION We have detennined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any signif-icant environmental impact. Having made this determination, we have further con-cluded that the amendment involves an action which'is insignificant from the stand-point of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environ-mental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.

/

Principal Contributors:

L. Frank, Materials Engineeying Branch, DE.

J. Rajan, Mechanical Engipeering Branch, DE J. Ridgley, Auxiliary Systems Branch, DSI R. Serbu, Radiological / Assessment Branch, DSI R. Biritel, LicensingeBranch No. 4,- DL

'Date:

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