ML18283B722

Letter Submitting Amendment No. 20 and 17 to Facility Licenses No. DPR-033 and DPR-52 for Browns Ferry Nuclear Plant, Units No. 1 and 2 in Response to 11/05/1975 Request

ML18283B722
Person / Time
Site:	Browns Ferry
Issue date:	03/03/1976
From:	Purple R Office of Nuclear Reactor Regulation
To:	Watson J Tennessee Valley Authority
References
Download: ML18283B722 (48)

Text

ep3 F6 Docket Nqs. 50-259 and 50-260 Gentlemen:

DISTRIBUTION'ocket NRC PDR Local PDR ORB-3 Reading KRGoller/TJCarter

.SMSheppard TVWambach OELD OIQE (7)

Tennessee Valley Authority BJones (8)

ATTN: -Mr. James E. Watson BScharf (15) hfanager of Power JMcGough 818 Power Building JSaltzman Chattanooga, Tennessee 37201

CMiles, OPA TBAbernathy, DTIE JRBuchanan, NSIC ACRS (16)

The Commission has issued the enclosed Amendments No. 20 and 17 to Facility Licenses No. DPR-33 and DPR-52 for the Browns Perry Nuclear Plant, Units 1 and 2.

These amendments are in response to your request of November 5, 1975, as supplemented November 28, 1975 and February 5, 1976.

These amendments authorize modification to Browns Ferry Nuclear'lant, Units 1 and 2 by approving the drilling of the fuel assembly lower tie plates of Types 2 and 3 fuel assemblies to provide bypass flow.

This bypass flow was originally provided for by holes in the lower core support plate.

By Amendments 17 and 14 to Licenses DPR-33 and DPR-52 for Units 1 and 2, respectively, authorization was issued to plug the holes in the lower core support plate to eliminate significant in-core instrument tube vibrations.

These amendments do not authorize operation of Units 1 and 2 with the plugged core support plate and drilled fuel assemblies.

Operation with these modifications will not be authorized until a later safety evaluation is completed that addresses the effects on operation.

Copies of the Safety Evaluation and the Federal Register Notice are also enclosed.

Sincerely, Sinai sr~0d i,y B. A. Puryle Robert A. Purple, Chief Operating Reactors Branch ii Division of Operating React rs

Enclosures:

See'next page QM Zooil0/

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0 Tennessee Valley Authority O. hhrch. 3, 1976 cc w/enclosures.:

H. S. Sanger General Counsel 629 New Sprankle Building

'noxville, Tennessee 37919 Athens Public Library South and Forrest

Athens, Alabama 35611

'hfr. 1filliam E. Garner Route 4

Box 354 Scottsboro, Alabama 35768 h)r. Thomas Lee Hammons

Chairman, Limestone County Board'f Revenue

Athens, Alabama 35611 cc w/enclosures and incoming:

Ira L. htyers, H.D.

State Health Officer State Department of Public Health State Office Building montgomery, Alabama 36104

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, NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 I

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TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLAIVI'NIT 1 AMEND5KNT TO FACILITY OPERATING LICENSE Amendmen'~o.

20 License No. DPR-33 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated November 5, 1975, as supplemented November 28, 1975 and Februaxy. 5, 1976, complies with the standards and requirements of the Atomic Energy Act of

1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of 'the Act, and the rules and regulations of the Commission;-

C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering

.the health and safety of the public, and (ii) that such activities will be conducted in compliance with the,Commission"s regulations; ll D.

The issuance of this amendment will not be inimical to the common defense and security or, to the health and safety of the public; and E.

An environmental statement ox negative declaration need not be prepared in connection with the issuance of this

.amendment.

2.

Accordingly, the license is amended by adding paragraph 2.C(6) to read as follows:

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.2.C.t,'6)

The facilitymay be modified by drilling bypass flow holes in Type 2 and Type 3 fuel assemblies as described in NED0-21091, "Browns Ferry Nuclear Plant, Units 1 5 2 Safety Analysis Report for

'lang Modifications to Eliminate Significant In-Core. Vibrations: and NEDE-21156, "Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibrations", 'dated. January 1976.

This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COhHISSION Date of Issuance:

March 3, 1976 Robert A. Purple, Ch ef Operating Reactors Branch Pl Division of. Operating Reactors

1 UNITED STATES t

NUCLEAR REGULATORY COMMISSION.

WASHINGTON, D. C. 20555

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'TENNESSEE 'VALLEY AUTHORITY DOCKET NO. 50-260 BRONNS FERRY NUCLEAR'PLANT'UNIT 2 AMENDMENT'TO FACILITY OPERATING LICENSE Amendment No.

17

.License No. DPR-52 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated November 5, 1975, as supplemented November 28, 1975 and February 5, 1976, complies with the standards and requirements of the Atomic Energy Act of

1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application,

. the provisions of the Act, and the rules and regulations of

.the Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activxties will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

An environmental statement or negative declaration need not be prepared in connection with the issuance of this, amendment.

2.

Accordingly, the license is amended by adding paragraph 2.C(6) to read as follows:

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0 2.C (6)

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The facility may be 'modified by drilling bypass floi holes in Type 2 and Type 3 fuel assemblies as described in NED0-21091, "Browns Fer'ry Nuclear Plant, Units 1 5 2 Safety Analysis Report for

'Plant. Modifications to Eliminate Significant In-Core Vibrations: and NEDE-21156, "Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibrations", dated January 1976.

3.

This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Robert A. Purple, Ch f Operating Reactors Branch ¹1 Division of Operating Reactors Date of Issuance:,

March 3, 1976

1

SAFETY EVALUATION REPORT ON THE REACTOR MODIFICATION TO ELIMINATE SIGNIFICANT IN-CORE VIBRATION IN BROÃNS FERRY UNIT l AND UNIT 2 DOCKET NOS. 50-259 AND 50-260

'By

'Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission

Table of Contents

~Pa e

1.0 Introduction...........................................

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2. 0 ackground B

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3. 0 Fuel Channel and Reactor Internals Inspection 3 ~ 1 inspection and Near Criteria......................

5 3 ~ 2 In-Core Instrument Noise.............

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4. 0 Evaluation of Reactor Changes

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4. 1 Mechanical Effects................................

13 4 ~ 2 Nuclear Performance and Thermal Hydraulic Effects

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16 5 ~ 0 Demonstration Tests.................................'...

17 5 ' 1 Mechanical

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5. 2 Thermal and Hydraulic...

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6 ~ 0 Post Reactor Modification Survei 1 1 ance

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1 TIPs 6'

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. 3 Internals

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24 6.

- 7 ~ 0 Environment a 1 Considerations

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.'8.0 Cone 1us 1ons

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1.0 Introduction By letter dated November 5, 1975, Tennessee Valley Authority (TVA) applied for amendments to Operating Licenses DPR-33 and DPR-52 for Browns Ferry Nuclear Plant Unit 1 and Unit 2 to authorize plugging of the bypass flow holes in the lower core support plate and drilling new bypass, flow holes in the fuel assembly lower tie plate.

In support of the application, TVA provided the General Electric report NEDC-21091, "Browns Ferry Nuclear Plant Units 1 and 2 Safety Analysis Report for Plant Modifications to Eliminate Significant In-Core Vibrations".

On November 14, 1975 Amendment 17 to DPR-33 and Amendment 14 to DPR-52 authorized the plugging of bypass flow holes in the lower core support plate.

By letter dated November 28,

1975, TVA submitted a non-proprietary version of the above GE report, NED0-21091.

This safety evaluation addresses the acceptability of drilling the fuel assemblies, but does not address the acceptability of reactor operation with the drill assemblies. 'he consid'eration of operation with the drilled fuel'ssemblies and plugged lower core support plate, along with any associated operating limits, will be the subject of a later safety evaluation report that must be completed prior to issuing, amendments that will authorize. such operation.

2. 0 Backgxound, 1n late

1970, a for eign 311R observed a change in the characteris-

'I ties of the readings from certain of the in-core instruments.

Sub-sequent examination of the fuel bundle channel boxes in the foreign reactor revealed significant wear on the corners of channel boxes adjacent to instrument. and source tubes.

This wear had led to crack-ing and holes in the channel boxes adjacent to the instrument that had displayed the anomalous readings.

The General Electric Company notified the NRC immediately of a possibly similar problem in domestic B>1R/0 plants.

Subsequently, the NRC ordered all the utilities with a similar reactor to inspect for this characteristic noise"" and.to notify the NRC if the noise level exceeded the prede-termined acceptable level.

The channel degradation was caused by vibration of instrument and source tubes excited by high velocity coolant flow from the 1-inch diameter bypass holes in the core support plate.

The presence of cracks or holes in a channel box is of concern since it would allow part of the cooling water that normally flows through the fuel bundle to flow out of the cracks or holes and by-pass the fuel rods.

Such a change in flow pattern would decrease the safety margins for the thermal per formance of the fuel.

These

'educed margins could lead to overheating and damage to the fuel in the event of some anticipated operating transients or some postulated accidents.

Significant wear and cracking of the channel boxes would also affect their mechanical strength for transients and accidents.

<<Noise is defined as the ratio of fluctuations in the signal in the frequency range of interest (generally 1-0 Hz)g divided by the mean value of the signal.

If large cracks occur in channel

boxes, there could be a potential for direct impacting of the tubes on fuel rods or interference with control rod movement.

The NRC ordered those plants with a high Traversing In-Core Probe (TIP) noise level to lower coolant flow and power to minimize the damage to the channel.

On July 18,

1975, the staff issued a

(1)>>

safety evaluation report stating that no further damage to the channel boxes is expected when the flow rate is reduced.

Also, the staff concluded that when the reactors are operated at the reduced power level described in the GE submittal the reactors will.

(2) not present an undue risk to the health and safety of the public,,

even with degraded channel boxes.

Some utilities, e.g., operators of the Duane Arnold and the Vermont Yankee BLAIR/4's, decided to shut down the reactors and plug the bypass holes in the lower core plate.

The NRC approved such an action and stated that'lugging only could result in an allowable power penalty for some reactors.

Concurrent with this action, GE has developed 'a permanent reactor modification to eliminate sign3.ficant in-core vibration.

The permanent modification consists of both drilling two holes 'in each fuel bundle lower tie plate to provide an alternate bypass flow path and at the same time plugging the 1-inch bypass holes.'he GE development of this permanent modification for the channel box wear problem has C

(6) been completed and reported to the staff The staff has completed

<<References are numbered and listed in Section 8.

its generic review of the permanent modification only for reactors em-ploying fuel bundles with the holes drilled in all lower tie plates in conjunction with plugging of all the 1-inch bypass holes (e.g.,

Browns Ferry 3).

The review is summarized in this safety evaluation report.

Concurrently the staff has reviewed the effects of drilling holes in the lower tie plates for some but not all of the fuel bundles within the core (e.g.,

Browns Ferry 1 and 2).

Since the number of bundles with holes drilled in the lower tie plate directly affects the bypass-region-t.o-bundle flow rates and the reflood rate for ECCS, the safety analysis for those reactors not having holes drilled in all fuel bundles must be reviewed on an individual basis.

Thus, the complete evaluation for operating limits on any reactor having

.drilled holes in only some of the fuel bundles is excluded from the scope of this summary.

However, the mechanical and hydraulic con-siderations of operating with only some of the fuel bundles having drilled holes were considered.

>.0.

Fuel Channel and Reactor Internal Ins ection 3.l Ins ections and Near Criteria As a routine part of planned reactor shutdowns, the chan-nel boxes and instrument and source tubes are visually inspected for corner wear.

.Cracks or holes in the channel 'boxes are readily apparent in the spent fuel pool without optical aids.

The results on each channel are compared with predetermined acceptance criteria for reuse.

The bases for establishing acceptable wear limits as well as the inspection plan are discussed in the GE report NEDC-20994 (4)

The radial depth of the wear on the channel box corners was estimated from an inspection procedure used at several Bi1R/4 reactor sites.

The inspection station was located at the fuel storage pool using a fuel preparation

machine, a borescope and a visual standard.

The channel wear was observable visually by the contrast between the Zircaloy-4 metal and the zirconium oxide adhering to the unworn por-tion of the channel box.

Cracks 'and penetrations were observable by their lack of light reflection.

The widths of the wear marks were measured by direct comparison with the known dimensions on the visual standard.

The depth of wear was inferred from a simple Pythagorean derivation for the radial overlap of two eccentric circles (Figure 2-2, reference 4).

This inference assumes no horizontal wiping of the tube on the channel.

The depth from uniquely wiping wear is only 42$

of that inferred by this technique.

Thus, the technique used to estimate coiner wear wa" conservative.

1 General Electric has performed visual inspections specifically for channel box wear at 18 reactors (9 with bypass flow holes in the lower core plate and 9 without bypass holes).

The results of all the reported inspections have been reviewed in detail by the staff.

More than 1600 channel boxes were examined during these inspections at those plants with bypass flow holes.

Only some in-core tubes are adjacent to bypass holes.

No unusual wear was observed at any chan-nel box corner not adjacent to in-core instrument or source tube.

The reject rate for channels adjacent to bypass holes is about two times higher than the reject rate for channels adjacent positions with no bypass holes.

Thus, the staff has.concluded that the joint presence of both in-core instrument and source tubes ard bypass flow holes was necessary to cause.significant channel box corner wear.

The results of the more detailed inspections at nine other reactors

, having no bypass holes in. the core plate have also been reviewed.

The inspections were focused upon more than 100 channels adjacent to in-core

'instrument and source tubes.

The results show that reactors without bypass holes in the lower core. support plate have exhibited no signifi-cant channel box corner wear.

General Electric recommends two types of channel inspections:

diagnostic and general.

The procedure is to diagnose the extent of wear by sampling selected channels and by performing a general in-spection for all the channels adjacent.to an in-core instrument tube only when the diagnosis yields evidence of significant wear.

>Ihen the channel wear problem was first identified GE, re-investi-gated their channel box design margins.

They found that when a chan-nel box corner was worn less than

.01 to

~ 02 inches (the nominal wall thickness is 0.08 inches) the original design limits were not violated.

This conclusion

>>as based upon a stress analysis of the channel boxes considering all modes of loading conditions such as steady state,

fatigue, steam line break and seismic GE identified fatigue as the limiting,design loading.

The fatigue loadings result from pressure variations from normal operations (e.g., startups and shutdowns, daily and weekly load reductions, and rod worth tests) as well as the various abnormal transients (e.g.,

pump trip, turbine trip, generator load rejection, etc.).

The ink'ormation supplie'd was not sufficiently

. (4) comprehensive to perform an exhaustive review of the channel integrity.

However, the staff performed several bounding calculations for maximum E

allowable wear and found that GE wear limits are acceptable.

There are four types of instrument and source tubes in a BNR.

They are Local Power Range Monitor (LPRM), Source, Intermediate Range Monitor (IRM), and Source Range Monitor (SRM).

When there is excessive vibration, these stainless steel tubes impact or rub against the Zircaloy channel box corners.

GE has inspected over half of the total number of in-core instrument tubes at two different BNR/4 reactors.

Two LPRM tubes were replaced because they exceeded GE's wear limits. It should be noted though that those two tubes were located where channels experienced through-wall wear and some pieces of the channel were tom off.

The GE criterion for allowable wear on the instrument tube is r

s approximately 20~~ of the nominal thickness and could mean that the tube resistance to collapse was reduced to half its original resis-tance.

The staff's calculation indicated that.Ol inches of wear does not constitute a significant reduction from the original safety margin.

Ne therefore conclude that the allowable wear for the SRM and IRM tube should not exceed 0.01 inches and the criterion be applied in all future plant inspections.

Furthermore, we require that all the in-core tubes be inspected prior to restart when the diagnostic inspection indicates that there is significant wear on the channels in a BNR/4.

3.2 In-Core Instrument Noise Mhen the core flow exceeds about 40 percent of rated flow for BMR/4's with bypass flow holes, the signal from the fission detectors of the LPRH subsystem and the TIP subsystem exhibit a characteris-tic noise associated with vibrating LPRt4 'instrument tubes.

This characteristic noise in the TIP traces and LPRtf time traces has a

frequency range of about 1 to 4 Hz.

However, other low frequency noise is also observed in these signals and is similar to that ob-served in BIB/3's.

The neutronic signals generated by the fixed LPRth detectors and the moveable (or parked)

TIP detectors and as recorded by plant or. special recording instrumentation can be correlated with the im-pacting of channel box corners and instrument tubes in a 'number of ways.

A direct approach consists of estimating the 1 to 4 Hz noise content in a TIP trace or an LPRt1 time trace; Another approach consists of'sing noise analysis techniques and computing either the power spectral density (PSD) as a function of frequency for a detector or the cross power spectral density (CPSD) as a function of frequency for a'y two detectors.

The acoustic>>'oise caused by impacting in-strument or source tubes on channel boxes can also be measured with accelerometers attached to instrument/source tube components that are external to the reactor pressure vessel.

Other approaches which use piezoelectric affects (TIP detector as a sensor) may also be used as an indicator of vibration.

<<The signals recorded with the accelerometers are termed "acoustic noise" in this report for the sake of brevity and convenience.

All of the various methods of relating observations on this impacting and vibration of instrument/source tubes indicate the same trends.

BliRs with plugged bypass flow holes in the lower core support plate indicate little neutronic or acoustic noise characteristic of the vibrating or impacting of instrument tubes on channel box corner s.

BiiR/4s with bypass flow holes open but with core flows restricted to 40 percent or less of rated flow also in-dicate similar results; But BllR/4s with bypass flow holes open and operating in the range of 40 to i00 percent of rated flow exhibit neutronic/acoustic noise varying from slight to considerable for the affected instrument/source tubes.

The measured channel box corner wear for several B>1R/4's has been shown to correlate with neutronic noise, either directly esti-mated or computed PSDs 'or CPSDs.

However, the correlations are not strong.

All that can be said is that the greater the neutronic noise with a frequency content of 1 to 4 Hz at a given location the greater the expectation.of channel box corner wear.

Establishing a reliable correlation is difficult due to the complexity of the phenomena (e.g.,

number and placement of bypass flow holes around an instru-ment source

tube, the motion of the affected tube and fuel channels, the control rod position and previous history, the in-channel void

content, the bypass region void content, core wide flux gradients, microphonic noise of the detectors, variations in core flow, and the quality of the plant measuring systems).

Quantitative aspects

T

of the effect of position and voids on the detector signal have been studied by our consultants at the Brookhaven National Lab-oratory The. calculations performed by our consultants (5) generally support the previously stated observations concerning neutronic noise caused by vibrating instrument tubes.

Although the effect of instrument tube movement and channel box corner wear on neutronic noise is generally understood, it is currently not possible to predict the occurrence of holes, splits, or cracks in channel boxes.

>le believe that the general complexity of the associated phenomena, the range of reactor opera-ting states and the lack of sophistication of plant instrumentation precludes exact predictions of the occurrence of holes, splits, or cracks in channel boxes.

However, we conclude that trends in measurements over a period of time, with reactor operation at substantial'core flow rates permits an assessment'f the po-tential for substantial channel box damage.

Therefore, based on our own analysis of the channel box corner wear data and neutronic noise, the study per formed by our consul-

tants, and a review of the information from domest'ic B0/R/As con-cerning channel wear and noise, we conclude that:

( 1)

BNRs with plugged or no bypass flow holes in the

'ower core support plate do not have any significant

neutronic or acoustic noise of the type associated with the channel wear problem, (2)

BMR/0s with bypass flow holes do not have any significant neutronic or acoustic noise, of the type associated with the channel wear problem, if the core flow is restricted to about

>IO percent of rated flow or less, (3) the measured neutronic and acoustic noise, for B'dRAs with bypass flow holes

open, increase as a

function of increased core flow, (0) neither neutronic or acoustic methods are presently capable of indicating the occurrence of holes,

splits, or cracks in a channel

box, and (5) noise measurements need to be evaluated over a period of time to monitor any changes or abnormalities as an indication of potential for'hannel box wear.

4;0 Evaluation of Reactor Chan es 4.1 Mechanical Effects General Electric has proposed to reduce the vibration of instrument and source tubes by eliminating adverse crossflow because of the 1-inch bypass holes in the lower core support plate adjacent to these tubes.

The design change proposed to eliminate adverse coolant crossflow at in-core tube elevations is to both drill two holes in each fuel bundle lower tie plate and to plug the bypass holes in the lower core support plate.

The two drilled holes are always located at the narrow-narrow interchanncl gap and not at the wide-aide gap where the flow might impinge on the control blades.

With all the bundles drilled there are approximately ten times as many holes as there were in the core support plate, and the total flow area 'is slightly less.

The holes in the fuel bundle lower tie plate are slanted to direct coolant flow down toward the-core support plate prior to mixing into the total bypass flow which is upwai d.

This results in a more uniform flaw throughout the core at elevations adjacent to the in-core tubes.

The uniformity of flow was demonstrated at the GE cold flow test facility by measuring axial velocity distributions.

Drilling only some of the fuel bundles is expected to provide a partial benefit of reduced adverse crossflow at elevations ad-jacent to in-core tubes.

Thus, no adverse effect on channel box wear is expected when operating with only some of th" bundles having holes drilled in their lower tie plates.

The lower tie plate serves to support the weight of the fuel bundle and rests on a fuel suppor t casting (see Figure 5-3, refer-ence 4).

Both components are. stainless steel.

The thickness of the tie plate wall is approximately 1/2 inch at the holes.

A stress analysis (including the stress concentration factor for the holes) indicated that the stress" levels are an order of magnitude below the allowable stress when all the expected loads are considered for

normal, abnormal and postulated accident conditions.

GE also investigated implications of a misoriented bundle where the flow would be directed toward the control blade.

Simu-lated tests in the cold flow facility at San Jose showed no abnormal conti ol rod vibration.

GE further examined the effect of this design change on other internal components (e.g.,

core support plate, guide

tubes, shroud support) and found no significant effect.

Plugging the bypass holes is also a part of the reactor modi-fication.

The staff's safety evaluation of such plugs was performed

'(

prior to issuance of the license amendments on November 14, 197S that

. authorized plugging of the bypass flow holes in the lower core 'suppoxt plates.

The conclusions of that evaluation are supported by the service experience of plugs at the Vermont Yankee and the Pilgrim 1 reactors where plugs were installed to eliminate control curtain vibration.

Post-service examination of an extracted plug exhibited neither degxada-tion nor wear of the plug after one fuel cycle.

The possibility of plug vibration from the flow through the two drilled tie plate holes was investigated by GE at the same cold test facility with full size plugs and tie plates.

No unacceptable plug vibrational response was found as measured by accelerometers.

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Long-term fatigue, creep and relaxation of parts of the plug

however, should be monitored by reasonable sampling inspection at each outage of the lead plants including some non-destructive and

(~)

destructive tests.

GE proposed an extensive plug surveillance program which the staff considers mandatory (see section 6).

Vhile developing and demonstrating the plant modification to eliminate wear caused by in-core tube vibration, GE has also developed a method of machining the lower tie plates.

The imple-mentation will be performed in two steps:

. drilling and deburr ing of the fuel bundle lower tie plate.

These operations on exposed fuel will be performed in the fuel storage pool under ab'out 25

,feet of water, s and The implementation procedure employs pneumatic drzll clamping devices.

Care has been taken in the design of the equip-ment to preclude misorientation of the fuel bundle.

The ver ifi-cation that all debris can be removed was demonstrated in a full-scale underwater test facility.

Ne observed the underwater machining pro-ceduree.

The rigors of the underwater machining procedure will necessitate close adherence by the personnel doing the machining to the specific Quality Assurance requirements.

General Electric has established several levels of contingency plans for possible difficulties during implementation.

The plans begin with simple procedures and progress to the replacement of the entire fuel bundle.

All contingency plans will be demonstrated before their implementation.

4.2'uclear'Performance and"Thermal'Hydr'aulic'Effects Since only some of the fuel bundles are being drilled, we will require prior to issuing amendments authorizing operation with the drilled fuel assemblies that either (1) a plant specific evaluation be submitted for a partially modified reactor, or (2) the plant nuclear and thermal hydraulic parameters, characteristics, and performance for normal, transient and accident conditions be based on the more conservative plugged-only core configuration (e.'g., reference 3).

5.0 Demonstration Tests GE per formed a cold hydraulic test at its San Jose facility to first determine the cause of in-core instrument tube vibration and channel box damage and secondly, to see that their pt oposed modifica-tions will perform satisfactorily as expected.

Thirty-two fuel bundles (4x8 array) were installed in a test tank with as-manufactured channel

boxes, lower tie plates, control rod plates, fuel support castings and in-core instrument tubes.

Plan views are given on pages 5-64 through 5-86 in reference 4,

There are some differences between the test and an in-reactor configuration.

The LPRH tubes in the test are cut short to approxi-

~ mately 15'nd attached to a spring whereas these tubes are more than 40'ong in-reactor,,All the internals in the LPRM tubes (TIP tube, fission chamber and cables) were" removed to facilitate installing an accelerometer.

The flow orifices of the fuel support castings were slightly altered to simulate the bypass flow volume.

In some tests, fuel rods were removed from the channels and replaced by dummy weights.

Also, the top of the fuel bundle is sealed (due to limited pump capac-ity) to simulate only bypass region flow and not flow through the fuel.

5.1 Mechanical.

For the initialBUR/4 simulation, GE was able to produce P

significant impacting of an LPRM tube and channel box.

>/hen the pro-posed modification for operating reactors was tested, the impacting l'evel was considerably reduced.

The staff monitor ed these tests and observed them on several differ'ent occasions.

Additional tests were performed at the Moss Landing facility.(6)

The test facility consisted of sixteen fuel bundles (4x4 array),

one 0.750 inch OD LPRH tube, four control rod blades, a shroud and a pressure vessel.

It:simulated in-reactor temperatures and pres-sures but no two phase flow was introduced.

Two conclusions were drawn from the tests.

First, the amount of bypass flow measured was more than expected.

Secondly, the impact level between fuel bundle and LPRH tube was higher than the value observed in the previous cold tests at; San Jose.

GE

, reduced the lower tie plate hole size from the original to correct for the desired bypass flow.

The reasons for the higher "g" level observed by the acceler-ometei in the LPRH tube were also investigated.

The difference can be attributed to.the in-bundle flow.

Xn the cold test, in-bundle flow was sealed'ff because of a limited pump capacity thus only'imulating bypass flow between channels.

When the flow was allowed to pass through the fuel bundle in a channel box at Moss Landing it caused a slight excitation of the fuel bundle thus adding to the LPRH tube vibration add impact.

GE confirmed bundle vibration at the cold facility by opening the flow seal to four fuel bundles.

Further tests were per formed at Moss Landing for both the BWR/3 simulated configuration and the fully plugged BWR/4 mockup.

GE found that the impact levels are. the same as that of the BWR/4 with the complete modification (ranging be-tween 4 to 8 g's).

They also confirmed, at the same facility, that the BWR/4 with bypass fl'ow holes in the core support plate produced accelerations about an order of magnitude higher.

GE concluded that since the impacts for the Bi!R/3 and for the modi-fied BWR/4 were equivalent and since no significant wear was observed in the BWR/3 channel inspections after full service life, the proposed BWR/4 modifications should eliminate the significant wear.

The Hoss Landing tests employed those core components for use in both the BWR/3's and the BWR/4 s (both modified and unmodi-fied).

Although the scale of the entire core was not simulated in the tests, the relative effects for the hydraulic and mechanical responses of the components were measured at Hoss Landing.

The measured impactings for tests from both th BWR/3 components and the modified BWR/4 components were significantly. improved relative to those from the unmodified SNR/4 components.

Based upon the above observations and the assumption that the outreactor tests are a

'caled equivalent of reactor hydraulic and mechanical environments, we conclude that the instrument and source tube impact levels in the modified BWR/4 s are expected to.be equivalent to the BWR/3's.

General Electric reported data to show that no significant wear from impacting has been observed in their BWR/3 surveillance program.

I 0

To provide verification of the expectations on actual operating'eactors, we believe that a comprehensive surveillance program is needed which is further discussed in section 6.

Final confirmation of the modification can only occur after the alternative flow path configuration has experienced a full fuel cycle of service.

The plants employing this modified configuration need to schedule a

post-irradiation surveillance on the channels at each outage for that purpose (see section 6).

5.2 Thermal and H draulic Alter nate flow paths and finger spring flow tests were performed by General Electric in the ATLAS facility which simulated the inlet geometry and bypass region for one fuel bundle under typical B)1R operating conditions.

GE has stated that all components used in these tests were typical of those in production and cur rently operating in BUR's which incorporate finger springs in the fuel design.

The test results provided the applicant with flow loss coefficients for different hole sizes and leakage flow r ates around the finger springs.

General Electric used these test results to determine the hole size to be drilled in the fuel. bundle lower tie plates.

6..0 P st Reactor Modification Surveillance In the previous sections we have discussed the necessity of having a surveillance program during reactor operation to guard against the possible recurrence of channel box degradation.

>le believe that two different types of sensors can be used to monitor vibrations during power operations:

( 1) in-core neutron detectors (TIPs),

and (2) accelerometers attached on the tube beneath the reactor which detects the mechanical energy of impact.

6. 1 TIPs Excessive instrument tube-channel box interaction pre-viously has been determined from the neutronic noise level in

~

'nfiltered TIP traces.

The plant modifications, including the plugging of the bypass flow holes, are expected to affect the noise content of the TIP traces.

In particular, the noise in the 1 to 0 Hz frequency range caused by vibration of instrument tubes should be reduced relative to power depe'ndent noise.

Based on our previous surveillance requirements, unfiltered TIP traces were taken prior to any plant modifications at the highest flow and power permitted.

For some plants, TIP traces were also taken at a number of power and flow conditions.

These h

data provide part of the basis for evaluating the efficacy of the reactor modifications.,

After the reactor modification, comparison of similar measurements with the pre-modification data will be made to confirm that the mechanical vibration of the instrument tubes has been substantially reduced.

The unfiltered TIP traces taken during return to power operation will also provide baseline data which can be used to monitor any changes in the 1 to 0

HZ noise level not attributable to such causes as power level, core flow and control rod pattern.

Therefore, we conclude that

( 1) surveillance using unfiltered TIP traces to monitor the efficacy of the plant modifications, and (2) the frequency of taking TIP traces in accordance with GE Standard Technical Specifications'(about 0 to 6 weeks of full power operation),>>

are an acceptable'mean for monitoring neutronic noise of the type associated with instrument tube vibrations.

, 6'2 Acc lerometer Since April 1975, when we first learned of in-core tube vibration, considerable experience has been accumulated both at various reactors and the San Jose facility regarding the capability of accelerometers to detect significant impact.

The Cooper, Duane Arnold and Peach Bottom reactors all demonstrated with acceleometers at different flow rates that there is a definitive transition in the flow rate below which no significant

<<GE STS Table 0.3. 1-1 Item 2e and footnote f (December 1,

1975 revision).

impact of the in-core tube can be detected.

This was the basis for allowing plants to operate at lower flow even though we suspected that some reduced wear rate may continue.

GE performed, an ex'periment with a full-length LPRH tube mounted upright in the air.

They then impacted the tubes with a hammer and monitored the stress wave with an accelerometer at various locations-along the tube.

NRC consultants and personnel from Philadelphia Electric Company, TVA and GE jointly experimented with a piezo-electric accelerometer at the Brown's Ferry plant during the current shutdown.

All came to the conclusion that the accelerometer is a viable sensor that detects any significant impact of the in-core tube.

The first two reactors to employ 'the modified configuration should install accelerometers on the in-core instrument tubes.

We regard this action necessary to provide further evidence of the efficacy of the modified reactor.

The applicants involved should establish a one month surveillance interval and report to us any anomalous behavior observed in the accelerometer.

GE has already accumulated some accelerometer experience in a BWR/3 plant.

This together with the experience obtained during power ascension flow tests at the Duane Arnold reactor and other (3) reactors with plugs only provides a reference for comparison.

6.3 Internals GE presented a plan to inspect channel boxes at the earliest refueling outage.

The first two reactors which imple-mented the plant modification will be required to perform detailed visual examinations of a statistically significant number of channel b'oxes for the first two refueling cycles after the modi-fication.

The results of current inspections indicate that outer pheripheral bundles may be more susceptable to a corner wear.

The statistical sampling should emphasize channel boxes which appear more susceptable to wear.

GE provided a satisfactory program for the plug surveillance.

It includes'removal of two plugs each from the core after

two, five and ten years of service, The plugs will be examined for wear, spring force relaxation and any deformation.

As discussed in section 2. 1, all the in-core instrument and source tubes should be inspected when the channel box inspection indicates that there is significant corner wear in the channels.

Furthermore, an in-core IRM or SRM tube must be replaced when its wear exceeds 0.01 inches.

- 2S-7.0 Environmental"Considerations Ne have determined that the amendments do not authorize a change in'ffluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.

Having made this determination, we have further concluded that the amendments involve an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 5S1.5(d)(4) that an environmental statement, negative declaration, or environmental impact appraisal need not be prepared in connection with the issuance of these amendments..

8.0 Conclusions Ne have reviewed the proposed reactor modification and found that:

(1) the outreactor flow test sufficiently demonstrated that the modification will reduce significantly in-core tube vibration and hence channel box damage; (2) the effects of the holes on the mechanical strength of the fuel assembly lower tie plate are insignificant; (3) the fuel rods and cladding of modified fuel will not be

~

damaged by the drilling operation; I

(4) measures to ensure that all drilling scraps and burrs are removed from the modified fuel are acceptable; and (5) the underwater drilling procedures satisfactorily protect the workers from radiation exposures.

1(e have concluded, based on the considerations discussed above,'hat:

(1) because the change does not involv'e a significant increase in the pxobability or consequences of accidents previously considered P

and does not involve a significant decrease in a safety margin, the change does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed modification to the fuel assembly, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.

The operation of Units 1 and 2 with the modified fuel assemblies will be the subject of later license amendments.

Date:

MAR 3 jg7g

- 28 References 1.

"Safety Evaluation Report for Partial Power Operation in Bl/Rs with Channel Llear " from V. Stello to K. Goller, July 18, 1975.

2.

Letter from R. Engle, GE, to V. Stello, NRC, July 11, 1975.

3.

"Safety Evaluation Report fov Duane Arnold Operation with Plugged Bypass Flow Holes",

from V. Stello to K. Goller, June 30, 1975.

4.

"Peach Bottom Atomic Power Station Units 2 and 3:

Safety An-alysis Report for Plant Modifications to 'Eliminate Significant

- In-core Vibration", NEDC-20994, GE, September 1975. (Proprietary) 5.

~A Study of the Effect of Position and Voids on B>/R In-Cove Detector Readings" by Hsiang-Shou

Cheng, BNL-20547, Sept.

1975.

6.

"Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibration", NEDE-21156, January 1976.

'Pr opr ietary) 7.

Georgia Power Company, Letter to gv.

A. Giambusso,

Director, Office of NRR, from I. S. Hitchell III, July 9, 1975.

8.

Hatch Unit 1 FSAR, Docket 50-321.

~

9.. Wilson, Grenda and Patterson, "The Velocity of Rising Steam in a Bubbling Two-Phase fiixture, ANS Transactions,'(1),

p.

151-'l52 (1962)

~

UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NOS.'0-259 AND 50-260 TENNESSEE VALLEY AUTHORITY NOTICE OF ISSUANCE OF AMNDMENTS TO FACILITY OPERATING LICENSES Notice is horeby given that the U. S. Nucloar Regulatory Commission (the Commission) has issued Amendment No. 20 to Facility Operating License No.

DPR-33 and Amendment No. 17 to Facility Operating Liconse No. DPR-52 issued to Tennessee Valley Authority (the licensee) for operation of tho Browns Ferry Nuclear Plant, Units 1 and 2, located in Limostone County, Alabama.

Tho amendments are effective as of the date of issuanco.

Those amendments authorize modification to Browns Ferry Nuclear b

Plant, Units 1 and 2 by approving the drilling of the fuel assembly lower tio plates of Typos 2 and 3 fuel assemblies to provide bypass flow.

This bypass flow was originally provided for by holes in the lower coro support plate.

By Amondments 17 and 14 to Licenses DPQ-33 and DPR-52 for Units 1 and 2, respectivoly, authorization was issued to plug the holes in the lowor core support, plate to eliminate significant in-core instrumont tube vibrations.

Those amendments do not authorize operation of Units 1 and 2 with tho plugged core support plato and drilled fuel assemblies.

Operation with theso modifications will not bo authorized until a later safoty evaluation is completed that addresses the effects on operation.

The application for these amendments complies with tho standards and requirements of, the Atomic Enorgy Act of 1954, as amended (tho Act),

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and tho Commission's rules and regulations.

The Commission has made appropriate findings as required by tho Act and tho Commissionls rulos and regulations in 10 CFR Chapter I, which are set forth in the license amendments.

Prior public notice of those amendments is not required since the amendments do not, involve a significant hazards considoration.

The Commission has determined that tho issuance of those amendments will not result in any significant environmental impact and that pursuant to'0 CFR 551.5(d)(4) an environmental statement,,

negative doclaration, or environmental impact appraisal nood not be propared in connection with issuance of'those amendments.

For further dotails with respect to this action, seo (1) the application for amendments dated November 5, 1975, as supplemented November 28, 1975 and February 5, 1976, (2) Amendment No. 20 to License No. DPR-33 and Amondmont No.

17 to License No. DPR-52, and (3) the Commission's related Safety Evaluation.

All of these items are available for public inspection at, t he Commission's Public. Document Room, 1717 H Street, N. N., Washington, D. C., and at the Athens Public Library, South and Forrest,

Athens, Alabama 35611.

A copy of items (2) and (3) may bo. obtained upon requost addressed to the U. S. Nuclear Rogulatory Commission, 1'Jashington, D. C.

20555, Attention:

Director, Division of Operating Reactors.

Dated at Bethesda, bfaryland, this day of FOR TIE NUCLEAR REGULATORY COtMISSION nal 8Igned Wj orrlca&

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Docket Nos.

50-259 and 50-260 Tonnessce Valloy Authority ATTN:

Hr. James E. Watson Hanagor of Power 818 Power Building Chattanooga, Tennessee 37201 January 9,

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Gentlemen:

You submitted Genoral Electric Company's'ropriotary report NEDC-21091, "Browns Ferry Nuclear Plant Units 1 and 2 Safety Analysis Report for Plant hhdifications to Eliminate Significant In-Coro Yibrations", with your lettor of November 5, 1975, and requested that tho xeport be with-

-hold from public disclosuro.

By lottox dated November 28, 1975, you submittod a non-propriotary edition of the report(HEM-21091).

The reason for withholding tho proprietary report (NEDC-21091) was stated to be that the information consists of the resuLts of analyses which have been made by GE at considerable expense and which represent significantly improvod analytical mothods.

Public disclosure of this information could enable knowledgoable competitors to qualify or modify their own design models to the detriment of the Genoral Electric Company's

~ competitivo position in tho industry.

Me have oxaminod the subject material and pursuant to Section 2.790{b) of 10 CFR Part 2, have approved your xequost.

Accordingly, pursuant to Section 2.790(b) of 10 CFR Part 2, wo aro with-holding tho proprietaxy report from public inspection.

Mthholding from-public inspection shall not affect the right, if any, of pexsons proporly and directly concornod to inspect the documents.

Sincerely, Original siCned by; RObeN*Pq I "

4g Robert A. Purple, Chief Opexating Peactors Branch Nl Division of Reactor Liconsing cc:

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H. S. Sanger General Counsel 629 New Sprankle Building Knoxville, Tennessee 37919 Athens Public Library South and Forrest

Athens, Alabama

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William E. Garner Route 4, Box 354 Scottsboro, Alabama 35768