ML20096B361

Safety Evaluation Supporting Amend 75 to License DPR-16

ML20096B361
Person / Time
Site:	Oyster Creek
Issue date:	08/27/1984
From:	Balukjian H, Brooks W, Voglewede J Office of Nuclear Reactor Regulation
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ML20096B356	List: ML20096B354 ML20096B361
References
NUDOCS 8409040116
Download: ML20096B361 (13)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 4-SUPPORTING AMENDMENT N0. 7s TO PROVISIONAL OPERATING LICENSE NO. OPR-16 GPU NUCLEAR CORPORATION AND JERSEY CENTRAL POWER & LIGHT COMPANY OYSTER CREEK NUCLEAR' GENERATING STATION J

DOCKET NO. 50-219 l

1.0 INTRODUCTION

By letter dated April 21, 1980 as supplemented March 9, 1981, August 31*, 1982, and October 28, 1983, GPU Nuclear Corporation (GPU) submitted a request for changes to the Oyster Creek Nuclear Generating Station's Provisional.

Operating License No. DPR-16 Technical Specification (TS) to accommodate the Cycle 10 reload.

The submittal included;NE00-24195, " General Electric 2

Reload Fuel Application for Oyster Creek" which provides the basis for the TS changes necessary for Core 10 operation. The ifcensee also provided submittals of July 22, 1983 which amended NEDO-24195, and May 1, 1984 which responded to the' staff's requests for additiona,1 information.

A Notice of Consideration of Issuance of Amendment to License and Proposed No Significant Hazards Consideration Determination and Opportunity for Hearing related to the requested action was published in the Federal e

Register on July 20, 1983 (48 FR 33081) and July 20, 1984 (49 FR 29495).

No request for hearing or public comments were received.

2.0 DISCUSSION AND EVALUATION-2.1 Methodology Topical j.

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2.1.1 Description of Report NEDO-24195 describes General Electric (GE) supplied reload fuel mechanic'ai design, nuclear evaluation met.:ods, steady state thermal-hydraulic methods, and reactor limits determination.

i In addition Appendix A presented the format in which the results

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of the reload analyses were submitted to the staff for review.

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previously approved [ Letter to Gridley (GE) from Eisenhut (NRC) dated May 12,1978] report NEDE-24011-PA, "GE Boiling Water Reactor Generic Reload Fuel Application," except for those j

features and considerations unique to the Oyster Creek reactor.

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. 2.1.2 Fuel Mechanical Desion Section 2 of NED0-24195, " Fuel Mechanical Design," describes the design bases, limits, and evaluation of the thermal, mechanical, and materials design of the Oyster Creek F.el System.. The fresh u

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fuel assemblies introduced in 'the Cycle 10 reload are General Electric Prepressurized 8x8 Retrofit (P8x8R) assemblies with lattice and bundle average enrichments described in Section 1 of NED0-24195.

As stated previously, the Reload Application for Oyster Creek (NE00-24195) generally follows the format used in the. General Electric Reload Fuel Application (NEDE-24011), which has been reviewed and approved by the NRC. Section 2 in both reports are identical except for the deletion of proprietary information from the Oyster Creek report. Because no significant differences exist.for the GE P8x8R fuel design -in the Oyster Creek application, the staff finds that previous approval of Section 2 in NEDE-24011 is equally applicable to the corresponding section of NED0-24195.

It should be noted that several amendments to NEDE-24011 are currently under staff review. Their-approval is expected to bring the most recent version of-NEDE-24011, new called the General Electric Standard Applicati6n for Reactor Fuel (GESTAR-II), into conformance with the NRC-Standard Review Plan, resolve several technical issues raised s_ince the original approval of NEDE-24011, and allow the application of a number of recently approved analytical methods.

In those areas where issues have been raised since the original approval of NEDE-24011 (e.g.,

fission gas release at high burnup), the Oyster Creek appli-cation has been modified (e.g., NED0-24195, Amendment 6, Section 5.5.2) in a ma ner similar to that used in the GE analysis of other operating reactors.

The staff finds this acceptable. ' Because the Oyster Creek Application (NEDO-24195) refers to the revision of NEDE-24011 which is approved by the date a specific analysis is initiated, final resolution of any technical issue of GESTAR-II will eventually (and automatically) apply to Oyster Creek. The staff finds this method of reference to future revisions of NEDE-24011 acceptable as well.

l 2.1.3 Nuclear Evaluation Methods q

Section 3 of NE00-24195, " Nuclear Evaluation Methods," describes the techniques used to obtain the r.uclear parameters of the fuel bundles and the core. Reference is cade to approved topical rt: ports which give detailed descriptions of the methods and their verification by comparison with measurements.

In addition the procedures used to determine the reference loading pattern for the new cycle are discussed, along with procedures for

" fine-tuning" the loading to account for unexpected events

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(e.g. end-of-cycle coastdownL in the current cycle). These procedures are-essentially identical to those in the approved

. generic reload design report, and are acceptable.

- 2.1.4 Steady-State Hydraulic Models.,

z Section'.4 of NF00-24195 " Steady-State Hydraulic Models,"

describes the methods used to obtain the pressure. drops and flow distribution for the steady-state therrial-hydraulic analyses of

. the core.;LThe present Exxon fuel bundles are modeled by using the characteristics.of General Electric bundles which are similar as regards to fuel rod geometry and enrichment. The core model.

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. requires hydraulic descriptions of orifices, lower tieplates, fuel rods, fuel rod spacers, upper tieplates,- the fuel channel.

and core bypass flow paths.'

The flow distribution to the fuel assemblies and. bypass flow paths is calculated on the assumption that the pressure drop l=

across all fuel assemblies'and bypass flow paths is the same.

This-has been confirmed by in-plant measurements as cited.by references. The components of bundle pressure. drop considered are friction, locale, elevation and-acceleration, each of.which.

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is discussed in separate subsections which include the

- applicable equaticns and references: The equations given for the various pressure drops were checkea and found to be in L

agreement with accepted practice.. A subsection on bypass flow indicates that full scale tests have been performed to compare against independent analytical models which predicted the test 4

results. One flow path where a significant amount of bypass flow can occur is due to channel wall deflection at the lower tieplate. To provide control.over this flow path, optional finger spring seals can be added to most reload fuel assemblies.

This can reduce bypass flow over a wide range of channel wall i

deflections. The analysis method provides for modeling fuel assemblies with and without finger seals by supplying separate hydraulic constants to represent both finger spring seal and non-finger spring seal bundles for proper calculation. Though

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the by-pass flow paths considered in the analyses are shown f]i -

schematically, no details are given to show that the analyses correctly model this flow. However, the steady-state hydraulic 2'

models use procedures which are essentially the same as those in the approved generic reload design report, NEDE-24011-P-A-1, and are therefore acceptable for use in the Oyster Creek reload fuel

!i application report.

2.1.5 Reactor Limits Determination i

The primary concern addressed in Section 5 of NEDO-24195 is to maintain' nucleate boiling on at least 99.9% of the fuel rods during normal operation and moderate frequency transient events.

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The measure cf transition from nucleate to film boiling used in BWR' reactor thermal' hydraulics is the Critical Power Ratio (CPR). A CPR of unity implies transition to film boiling; less than unity implies film bpiling which could lead to fuel damage. The CPR is dependent upon a numbe' of operating r

parameters such as core flow, core power, and reactor pressure.

These are subject to change during normal operation and have certain measurement uncertainties. Statistical analyses are conducted which take into account these uncertainties and operational variations to determine the CPR that meets the 99.9% criteria for normal operation; this value of CPR is called the safety limit minimum critical power ratio (safety limitMCPR). The safety limit MCPR protects the core during normal operation. However, a higher value of CPR, the operational limit of transient MCPR, must be maintained to

-prevent violation of the 99.9% requirement during transients of moderate frequency.

Section 5 of NEDO-24195 describes.. proc.edures for setting operating limits for the cycle and for setting certain limits (e.g., APRM rod blocks) iii the protection system.

These limits are established so.that violation of fuel thermal limits does not occur for normal operation or anticipated operational occurrences-and that acceptable consequences are calculated for acc.idents. Limits on the core-wide value of MCPR and clad strain are established.

The various anticipated transients are then examined for their effect on MCPR and linear heat generation rate. The change in these quantities during the transients is then combined with the safety limit to obtain required operating ifmits for each transient.

The most limiting of these limits then becomes the Technical Specification value for the parameter (e.g., MCPR).

l 2.1.5.1 Reactor Limits Determination Overview Section 5.0 presents an overview which addresses the determination of reactor operating limits required to meet safety requirements. The safety limit applicable to this review is that of selecting values of CPR which will provide an adequate safety factor during normal plant operations and during moderate frequency transients. Two separate entities a

are used in this report--both of which are called MCPR. The first is the safety limit MCPR which will ensure that 99.9%

of the fuel rods are in nucleate boiling. This value of CPR must exceed unity by a certain margin to take into account fluctuations in plant operating conditions and measurement and computational uncertainties in plant operating parameters.

This margin is determined'by conducting statistical analyses which model the plant and take' into account the range of l

. variation and treasurement uncertainties of the parameters which affect CPR. General Electric found that a value of safety limit MCPR of 1.07 is sufficient. Section 5.1 deals with determining this quantity. The second y.alue of MCPR of interest is the operating limit MCPR. The operating limit MCPR is the value of MCPR at which the plant must be operated to ensure that the MCPR will not fall below the safety limit MCPR during a transient.

This value is obtained by detemining the largest drop in critical power ratio found in any of the analyzed transients and adding that to the nomal MCPR. The analyses showed that Oyster Creek should be operated at a CPR at or above 1.25 with the General Electric reload. Section 5.2 discusses the operating limit MCPR.

Section 5.4 describes the analyses used to show that the reactor will cperate stably with the reload and Section 5.5 describes the accident evaluation methodology.

Section 5.0 is essentially identical to the staff approved report, NEDE-24011-P-A-1,'and is therefore acceptable.

2.1.6 Fuel Claddino Integrity Safety Limit

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Section5.1describeshowthenormahMCPRisobtained. This section is very similar to NEDE-24011-P-A-1 but contained several items requiring clarificat-ion or additional info'rmation. These include:

(a) the validity of the modeling of the Exxon bundles j

(which make up 80% of the reload core) by using i

fuel which most closely matches the thermal-hydraulic and nuclear characteristics of the Exxon fuel.

(b) the statement that the "large reload core analysis results conservatively apply to Oyster Creek for all General Electric-supplied reload cycles."

(c) that the General Electric bounding analysis is conservative although same of the plant-unique uncertainties may be girater for Oyster Creek.

In a'let'ter dated May 1, 1984 GPU provided additional information demonstrating that the Exxon bundles can be i

modeled by General Electric bundles. General Electric stated that NED0-24195 will be updated by including the results of an ODYN code analysis for Oyster Creek. This analysis uses a thermal-hydraulic medel for Exxon fuel and adequately resolves these issues.

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. 2.1.7 MCPR' Operating Limit Calculational Procedure Section 5.2 describes the procedure for obtaining the operating

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limit MCPR, the MCPR which wil1 keep 99.9% of the fuel rods in

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nucleate boiling during moderate frequency' transients. Topics which are discussed include:

(a) the system model used;

.(b) nuclear considerations (scram and void reactivity, Doppler coefficient); (c) inputs to transient ana;yses; (d) need to increase the operational limit MCPR at low flow; (e) transients chosen for_ analysis; (f) description of the analysis of each of '

those four transients; (g)_ exposure-dependent limits; and (h) effect of fuel'densification on MCPR operating limit.

This section closely parallels the same section found in NEDE-24011-P-A-1 and ~was acceptable after a few items were clarified. This included questions on whether or not several revisions to the subject report and several references to the reference report were NRC approved.

Also the licensee submitted a response related to.the margin between safety.

valve setpoints _and peak transient pressures and is acceptable.

2.1.8 Stability Analysis Method

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Section 5.4 is identical to NEDE-24011-P-A-1 except for two revisions dated September 1979 and one revision dated January 1980. These revisions we're found acceptable by the staff.

2.1.9 Accident Evaluation Methodology Various accidents are analyzed in Section 5.5 to determine the operating limits'(initial conditions) which preclude unacceptable consequences. For example, operating limits for the average planar heat generation rate (APLHGR) are established so that peak clad temperatures are not exceeded in the loss-of-coolant accident.

The following transients and accidents analyses are described.

Transients

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o turbine trip without bypass o generator load rejection without bypass o loss of feedwater heating o feedwater controller failure o rod withdrawal events 4

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o control rod drop o loss-of-coolant accident

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1 o fuel misleading o' recirculation pump sefsure-

- o refueling accident.(assembly drop)

-In addition,~the establishment of the core MCPR safety limit and the. analysis of core thermal-hydraulic stability are 4

discussed. The discussions are essentially.the same as.those in the approved generic top 1 cal.

2.2 Summary of Evaluation '

The following discussion summarizes the evaluation of NE00-24195 by the staff.

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2.2.1 Fuel Mechanical Design

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The' fuel-thermal,mechanicaland'materialsdesignmethods employed for Oyster Creek reloads are described in Section 2 of NED0-24195. Section 2 of that-report is essentially the same as the corresponding section.of NED0-24011, which has been reviewed and approved by the~NRC for reference in the safety analysis of other boiling water reactors.

Because the procedures employed in the reload design and analysis are i

identical (as far as the fuel. design is concerned), the staff finds the Oyster Creek application acceptable.

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2.2.2 Nuclear Design The nuclear design methods employed for Oyster Creek re' loads.

are described in the various topical reports referenced in Section 3 of the report. These reports have been reviewed l'

and approved by the staff for reference' to design methods for boiling water reactors. They are therefore acceptable for t

use for.0yster Creek..

i The procedures employed in the reload design and analysis are essentially the samt as those described in the previously approved NEDE-24011-P.and are acceptable. The procedures q

l used to' establish operating limits are similar to those

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previously approved and are acceptable.

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p 8-p 2.2.3 Thermal-Hydraulic Design The-thermal-hydraulic design methods employed for Oyster Creek ~

reloads are described in Sectjons 4 and 5 of NEDO-24195. These L

sections are essentially the same as the corresponding sections of NEDE-24011-P, which were previously reviewed and approved by the NRC for reference in the safety analysis of boiling water reactors. 'Since the procedures employed in the reload design

. and analysis are essentially the same, the staff finds the j.-

.0yster Creek application acceptable. Similarly, the procedures used to establish operating limits are similar to those previously' approved and are acceptable.

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2.2.4. -Reactor Limits Determination

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The a~ alyses.for-the rod drop accident, the fuel misloading event, and the control rod misoperation events have been reviewed as part of the Systematic Evaluation Program. 'These analyses have been. approved [ Approval letters from Crutchfield (NRC) to Finfrock (0yster Creek) dated March 31 and April 9,1981). -It was concluded that the analyses of l

these events meet present day requirements and criteria and f -

are acceptable.

2.3 LOCA Analysis

,. J The LOCA analysis for the Oyster Creek reactor is based upon approved 1-codes with two minor differences. The first difference is that the approved codes were written for jet pump configurations while Oyster Creek is a non-jet pump plant. Hence, the codes had to be modified to reflect this difference in plant design. The staff concludes that the i

a modifications made are acceptable. The second difference is that the approved codes allow for only two recirculation loops while the Oyster

,1 Creek plant has five loops.

In the Oyster Creek analyses the intact

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loops were combined into. a single loop and the broken loop was modeled 1

as the second loop. The combining of loops in computer simulations for-accidents analyses is standard industry practice and is acceptable.

It was. determined that data existed on the adequacy of the distribution of the low pressure core spray. The core spray is necessary to the recovery from a LOCA and there is, in 10 CFR 50, Appendix K, an allowable value for the convective heat transfer coefficient. The i

experimental data showed that the spray distribution was adequate up to j

55 psia of steam. However, there were no data for greater pressures

'and credit is taken for the initiation of core spray at 125 psia a

The licensee has provided information to show that, for pressures in excess of 40 psia, the heat transfer due to steam cooling (i.e., steam, produced by flashing and boiling, rising through the fuel bundles) is y

sufficient.to satisfy' the Appendix K requirements.

In addition, it is l'

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-shown that core. spray could be delayed to an initiation setpoint of 40 psia for breaks up to one square foot in area without exceeding a.

peak clad temperature of 2200 degrees Fahrenheit.

(It should be noted

.that the large break LOCA produces a.yery rapid depressurization and therefore uncertainty in spray distriEution above 55 psia is of no consequence.) This was shown by comparing the heat transfer coefficient from the fuel rods to the steam which was based on the Dittus-Boelter correlation..These coefficients were referenced downward to the saturation temperature so as to be able to apply them in a previously approved code. The steam flow rate was calculated by determining the amount of boiling in the active core and adding that to the amount of steam produced by flashing that is directed to the' active rods (conservatively shown to be' 50 percent of the total flashing).

The approved codes were then exerc'ised. The steam cooling heat transfer coef ficient remains above the spray coefficient of 1.5 for the period of concern; therefore use of the spray coefficient during that period is acceptable.

In sumary, the LOCA analysis for Oyster Creek is acceptable. The computer codes were modified in an acceptable manner in order to make them applicable to a non-fet pump plant.and-adequate core cooling will exi_st to bring the plant to a safe shutdown.

2.4 Evaluation Procedure The review of topical report NED0-24195 ha been conducted within the guidelines provided for analytical methods.in the Standard Review Plan (NUREG-0800, Section 4.3). Sufficient information is provided in this report and the referenced topical reports to permit the conclusion that the methods and procedures described are state-of-the-art and are acceptable.

2.5 Regulatory Position Based on the review of topical report NE00-24195 described above the i

. staff concludes that the report is suitable for reference by Oyster f

Creek in reload reports and other licensing actions to which it is applicable.

• i 2.6 TechnicalShecificationChangesandReload9(Cycle 10)

Changes Due to Methodology I

The staff has reviewed the proposed changes to the Technical Specifications which are intended to make them consistent with the new methodology.

Many of'the changes are editorial in nature - e.g., replacing the references to methods with NED0-24195. These are acceptable.

Others include:

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Replacing the' peaking factor ratio with' the ratio of the fraction-of rated power to the maximum fraction of linear power density in the APRM' scram and rod block functions in Section 2.3 and the bases.

2.- lhe acceptance criterion for the' hod drop accident has bien changed to make it consistent with the use of the Banked Position Withdrawal Sequence and the GE analysis method.

3. :The requirement for a control rod density of 3.5 percent nas been deleted since new analyses are performed with the.end-of-cycle, all-rods-out scram curve.

The staff finds the changes to be acceptable.

2.7 Cycle 10 Changes The Cycle 10 core consists of 172 fresh GE p8x8R fuel assemblies and 388 partly burned Exxon Type VB assemblies. The supplemental reload i

submittal contains a core loading diagram.

2.7.1 Fuel Mechanical Design 1

' The mechanical performance of the General Electric fuel-in the Oyster Creek Cycle 10 core has been-analyzed with the methods described in General Electric Reload Fuel Application for Oyster t

l Creek (NED0-24195). Our review o'f the GE Application for Oyster Creek is described in Section 2.1.2 of this report. The mechanical performance of the remaining Exxon fuel in the Cycle 10 core has been analyzed as part of the previous reload applications from this plant. Where the Exxon fuel affects the safety analysis of Cycle 10 specifically (this is largely.due to the thermal and hydraulic characteristics of the fuel), the analyses have been performed by General Electric using Exxon fuel characteristics provided in the Oyster Creek FDSAR (and Appendix B of NED0-24195). With regard to the fuel thermal, mechanical and materials design, the staff finds this application acceptable.

4 2.7.2 Nuclear Design The nuclear design and analysis of the core was performed with the methods described in NED0-24195. The staff has reviewed the results of the analyses.

The following comments are relevant.

o The effective multiplication factor of the core is less than 0.99 at cold (20*C) xenon-free condition with the strongest control-rod out of the core.

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.o The-standby liquid contro1 system is capable of producing a shutdown margin of 0.044 in the cold xenon-free state.

o The reactivity coefficients,are within the range'of those

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. usually seen for BWRs.

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'Because:the analysis was performed with acceptable methods and the results are acceptable, the staff concludes that the

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nuclear design and analysis of Cycle 10. is acceptable.

2.7.3 Thermal-Hydraulic Design The -thermal-hydraulic performance of-the General Electric fuel in the Oyster Creek Cycle 10 core has been analyzed with the methods described in NED0-24195 for which the licensee's application for Oyster Craek is described in Sections 2.1.4 and 2.1.5 of this report. The thermal-hydraulic performance of the remaining Exxon fuel in the Cycle 10 core has been analyzed

' as -part of the previous reload applications from this. plant.

General' Electric has developed a. thermal-hydraulic model for non-GE fuel (Exxon 8x8 Type VB) for-use in transient ' analyses with a mixed core of GE and Exxon fuel. The thermal-hydraulic nodel for the Exxon fuel design ~is based on the geometry of the Exxon fuel and pressure drop data-fer the Exxon fuel. Other thermal-hydraulic characteristics _of the Exxon fuel were assumed-to be identical to GE fuel.

In Kppendix B of NED0-24195 the geometry of the Exxon fuel is tabulated in Table B-1 and the 4

thermal-hydraulic model assumptions in Table B-2.

The transient results for Exxon fuel are in Appendix A of NEDO-24195 along with those-for the GE fuel. The staff finds the thermal-hydraulic design for this application to be accertable.

2.7.4 Transient and Accident Analysis The Rod Withdrawal Event - The rod withdrawal error analysis was P

performed with the methods described in NED0-24195.

If it is assumed that the limiting failures have occurred in the APRM rod block circuitry the change in CPR during the event is 0.27 for 1

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the GE fuel and 0.33 for the Exxon fuel. This is the limiting h

event for Cycle 10 and dictates the MCPR operating limit of 1.40.

1 Analyses have been performed for less than limiting failures, J

for which the change in CPR is smaller,' decreasing to 0.10 for no failures. However, no credit is taken for this fact in establishing cycle MCPR' operating limits.

Fuel Misloading Event - The effect of the misorientation of a fuel assembly in the core has been analyzed by the methods in NED0-24195.

The' limiting case results in a change in CPR of 0.20.

Thus, this event is not limiting for Cycle 10.

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1, Rod-Drop Accident. The consequences.of' the postulated -rod drop _

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event for Cycle 10 have been calculated by methods described in.

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.NE00-24195.' The Banked Posit.i.on Rod Withdrawal Sequence will be i

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employed by the licensee. This sequence ha's been shown to

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1. s inoperable rods make adherence to the programed sequence impossible, a rod worth check will be made to show that the maximum rod worth is-less than 1.0 percent ak. This value of rod worth has been shown to meet the 280 calories'per gram

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criterion. This_is an acceptable procedure and the staff concludes that an adequate analysis of the rod drop event has n

been performed.

2.7.5 Technical Specifications

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Supplemental information submitted on October 28, 1983 proposed to incorporate a new scram setting for. recirculation flow at 117

. percent rated flow. This change.is. conservative since it results

.;J in a new Technical Sprcification.(TS) requirement which did not previously exist in the current-TS and is also conservative 1,

relative to the previous setting of:120 percent which had been previously established as an administrative centrol contained in the plant procedures. The change.in MCPR limit from 1.3 to 1.4 -

4 is in the conservative direction and. furnishes the maximum allowable average planar LHGR curves for 5 and 4 loop operation.

Also,- the change revises the control rod withdrawal sequences and establishes the maximum in sequence' rod worth to be 1.0 percent ak. This change incorporates the use:of banked position withdrawal sequences which is more conservative than the previous -

withdrawal sequence. The staff has reviewed the-TS changes proposed for Cycle 10 and conclude that they are acceptable.

!i 3.0 ENhIRONMENTAL CONSIDERATION 1

This amendment involves a change in the installation or use of a facility

.j' component located within the restricted area as defined in 10 CFR Part 20.

,t The. staff has determined that the amendment involves no significant increase

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in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in

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individual or cumulative occupation radiation exposure. The Connission has 4

previously. issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public coment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).' Pursuant to t

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-10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

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4.0 CONCLUSION

The staff concludes that the Oyster Creek reactor may be operated for Cycle 10 without undue risk to the health and safety of the public. This conclusion is based on the fact that acceptable methods and procedures were' used to perform the design and_ analysis of the cycle and that the Technical Specifications have been correctly based on the results of that analysis.

The staff has also concluded, based on the considerations discussed above, that:

(1) there is-reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the h'ealth and safety of the public.

- 5.0 ACKNOWLEDGEMENT This evaluation was prepared by W. Broo'ks, J. Voglewede, and H. Balukjian.

Dated: August 27, 1984

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