ML19340A144

Suppl 3 to SER for Facility

ML19340A144
Person / Time
Site:	Oconee
Issue date:	01/29/1974
From:	US ATOMIC ENERGY COMMISSION (AEC)
To:
References
NUDOCS 8001100610
Download: ML19340A144 (25)

Text

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SUPPLEMENT NO. 3 TO THE I

SAFETY EVALUATION BY THE DIRECTORATE OF LICENSING U. S. ATOMIC ENERGY COSMISSION IN THE MATTER OF DUKE POWER COMPANY-OCONEE NUCLEAR STATION UNITS 2 AND 3 DOCKET NOS.- 50-270, -287 '

i JAN.2 9 1974

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1 TABLE OF CONTENTS

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1.0 INTRODUCTION

1-1 1.1 ' General 1-1 l

1.2 Scope of Review 1-3 2.0 MECHANICAL INTEGRITY OF CLADDING 1 3.0 EFFECTS OF DENSIFICATION ON STEADY STATE AND TRANSIENT OPERATION 3-1 3.1 General 3-1 3.2. Fuel Rod Thermal Analysis 3-3 3.3 Steady State and Loss-of-Flow Transient 3-4 3.4 -Other Transients 3-5 3.5 Conclusions 3-2 4.0 ACCIDENT ANALYSES 4-1 4.1 General-4-1 4.2 Locked Rotor Accident 4-3 4.2 LOCA Analysis 4-4 4.4 Rod Ejection Accident 4-6 5.0

SUMMARY

AND CONCLUSIONS 5-1

6.0 REFERENCES

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1.0 INTRODUCTION

'1.1 General Duke Power Company (the applicant) applied for an operating license for the Oconee Unit 3 reactor by application dated June 2, 1969.

The Atomic Energy Commission's Regulatory staff (the staff) subsequently completed its review of the application and issued a Safety Evaluation Report on July 6, 1973. A notice of intent.to issue an operating license was published in the Federal Register on August 10, 1972, by the Atomic Energy Co= mission. No hearing was requested.

On November 14, 1972, the Regulatory staff' issued a r. port entitled,

" Technical Report Densification of Light Water Reactors Fuels"(

which resulted from the staff's consideration of the Ginna fuel densification phenomenon. Based upon the findings in this report the staff requested

-on November 20, 1972 that the' applicant provide analyses and relevant bases, in accordance with the densification report,(1) that determine the effects of fuel densification on nor=al operation, transients and accidents for.the three Oconee Units. On January 16, 1973 the applicant filed a response to the request (2,3) for Oconee Unit 1 as a lead plant

.for this evaluation. On March 14, 1973, the staff requested additional information. The applicant filed a response to this request on April 13,-

~ 1973. (4,5)

On December 14, 1973 the applicant filed a response to these requests specifically for Oconee Unit 3.

• Numbers in () refer to references listed in Section 6.0.

• • The "Oconee 2 Fuel Densification Report," BAW-1395 (Proprietary) June 1973(11) was filed by the applicant in response to these requests specifically for Oconee.

Unit 2.'

The staff's evaluation is Supplement No. 1 to the July 6, 1973 Safety Evaluation Report.

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1-2 The "Oconite 3 Fuel Dencification Report", BAW-1399,I1N Nove=ber 1973, (Proprietary) provided an analysis of the effects of fuel densification using the most conservative values from the fuel specification since complete as-built data were_not available at the ti=e of the analysis.

The staff finds this acceptable and will verify that as-built data is within the fuel specification prior to finalizing the Technical Specification-Furthermore, since batch III fuel as described in BAW-1399 is to be

- resintered and sorted into three diameter groups,.the loading program will be reviewed for acceptability by the staff prior to finalizing the Technica1' Specifications.

The staff's technical review of fuel densification as it applies to Oconee Unit 3, and the technical evaluation of the applicant's safety analysis of steady state operation, operating transients and postultited accidents taking into account the effects of densification are presented in'this supplement.

This evaluation relies upon the July 6, 1973 Regulatory staff report " Technical. Report on Densification Report of Babcock & Wilcox Reactor. Fuels"( } which concluded that B&W's fuel densification models are in compliance with the staff's initial densification report (1),

The staff has concluded that the operation of Oconee Unit 3 for the first cycleLat power levels up to 100 percent'of full power, in accordance with the Technical Specifications, will not present an undue risk to the health and safety of the public.

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1-3 1.2 Scope of Review The essential ele =ents that cust be considered in evaluating the effects of fuel densification have been set forth in the staff's initial densification report. (

Since t.ie performance of the facility in steady state operation and during various postulated transients and accidents had been established previously as rpported in the_ Final Safety Analyses Report (FSAR) without the assu=ption of fuel densification, it was only necessary to evaluate those changes J

in the analyses and in the results that are attributed to fuel densification. The effects of fuel densification on the steady state operation and on the course of postulated plant transients and accidents were evaluated by the applicant and reviewed by the staff.

The staff reviewed the effects of fuel densification for Oconee Unit 3 using.the staff's guidelines, the technical evaluation of the applicant's safety analysis of steady state operation, operating transients and postulated accidents and the generic evaluation (0} of B&W methods for assessing fuel densification and its effects' In the evaluation the applicant appropriately considered the staff' guide--

lines including the effects of instantaneous and anisotropic densification (initial density minus 2a, and final density 96.5% TD),

and the assunption of an axial gap leading to a power spike. The staff reviewed the effects of fue1' manufacturing and reactor operating parameters on the fuel densification mechanism. The generic

' evaluation of these items is included in' Reference 6.

The staff reviewed B&W's assumptions, methods, and computer codes used in.

. evaluating the fuel densification effects. The generic evaluation

1-4 of B&W's c:odels is also included in Reference 6.

The mechanical integrity of the fuel cladding and the thermal state operation, operating transients, and postulated accidents are discussed in the following sections.

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2 2.0 MECHANICAL INTEGRITY OF CLGDING Clad creepdown during the core life is not considered by the applicant.in the calculation of gap conductance. This'is a conserva-tive assumption since the reduced gap size due to clad creepdown wou22 result in a higher gap conductance and thus in a lower stored energy 1n.the fuel. The staff reviewed the B&W cethod for calculating

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the clad collapse time (6), which is the time required for an unsupported cladding tube to flatten into the axial gap volu=e caused by fuel-densification. On the basis of independent staff calculations and

- from experience of fuel performance in.cher reactors, the staff concurred.with the applicant that clad collapse is not expected for the Oconee Unit 3 fuel during the first cycle of 10,944 effective full power hours (EFPH)..However, the staff concluded that the evaluation model for collapse time calculations contains several deficiencies in its application to Oconee Unit 3.

The staff infor=ed dhe applicant that an acceptable model for collapse time calcula-tions is necessary for subsequent fuel cycles of Oconee Unit 3.

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3-1.

3.0 EFFECTS OF DENSIFICATION ON STEADY STATE AND TRANSIENT OPERATION 3.1 General Fuel densification can affect the steady state operation because-of axial gaps in the' fuel coluen that results in local neutron flux spikes and an overall increased linear heat rate. An additional

.effect occurs in the transient analyses since, due to a lower gap 4

. conductance, the fuel has'a higher initial stored energy and a slower heat release rate during the transient.

On the basis of evaluations of the effects of fuel densification the Oconee Unit 3 reactor will be operated with more restrictive limits on control. rod patterns and position than originally proposed, and with a' reduced maximum linear heat generation rate. The changes. con-sider'the effects of local peaking caused by gaps in the fuel pellet

-stack and changes in the gross peaking factors, primarily axial, which can be achieved by more restrictive operation of control rods.

The effects of densification on power density distributions have

'been calculated using sodels'in conformance with those discussed in Section 4 of the. staff densification report.(

The primary calcula-tions used the models and numerical data of.the Westinghouse power spike model as described in Appendix E of that report, as appropriate l to' B&W fuel except that the initial nominal density used was -[]* and tF*

probability of gap size was changed to conform to that recommended eh eenff.(1)'

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• [:] Brackets denote data known by the staff and considered' proprietary to the' applicant.and specified in references 4 and 5 to this report.

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-The calculations by the applicant take into account the peaking due to a given g'ap; the probability distribution of the peaks due to the distribution of gaps, and the convolution of the peaking probability.

with the design radial power distribution. The calculations result in a power. spike factor that varies almost linearly with core height and reaches a maximum value of 1.1 at the top of the core. The overall calculation falls with'in the range' examined,(

} by our r.onsultant, Brookhaven National Laboratory, in conjunction with revir.ws of other models.

A normalized shape for the power spike factor is derived from power spikes caused by different. gap sizes at various axial locations.

The normalized shape is then _ used'in conjunction with various axial power shapes to determine the axial location.at which the decrease in-DNBR due to the superimposed power spike'is maximized. These cal-culations also include the increase in average linear heat generation rate from 5.66 Kw/f t to 5.82 Kw/f t due to' the reduced fuel column height. based on the instantaneous densification-from the minimum initial

' density of' [ :] theoretical density. (TD) to a final' density of -

}.The reactor operating limits, which will'be part of the

-0.965 TD.

--Technical Specifications for Oconee Uni:: 3, are based on. maximum

. linear heacLgeneration rate.through.the reactor. power vs axial-

offset ' correlation.

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3-3 3.2 Fuel Rod Thermal Analysis The' applicant uses the B&W computer code, TAFY( }, to calculate gap conductance, fuel temperature, and stored energy for the Oconee Unit 3 fuel, which in turn are used in the safety analyses. To 4

demonstrate the applicability of.the TAFY code for the evaluation of the Oconee Unit 3 fuel thermal behavior, the applicant co= pared TAFY predicted fuel te=peratures and gap conductance with experimental data.

The staff. reviewed the TAFY code and concluded that appropriate assu=ptions have been used for modeling of the physical phenomena incorporated into the code (thermal expansion, fuel swelling, sorbed gas release, fission gas release), with two excepticas:

(1) partial contact between the clad and fuel and (2) formation of,a central void due to fuel restructuring on the basis of columnar grain growth at a te=perature of 32000F. Details of the staff's evaluation of the TAFY code and its application to Oconee Unit 3 type fuel rods are given in' Reference 6.

I Because of the two exceptions noted above, the staff required the applicant to analyse the fuel thermal performance using a 25%-reduction in gap conductance and taking no credit for fuel restructuring. This analysis ('} resulted in a reduction in the peak linear heat rate at which centerline fuel melting would occur from 22.2 Kw/ft before densi-fication to 20.15 Kw/ft after densificatio.. was conservatively taken f

d 3-4 into account. The reactor protection system prevents fuel centerline

- melting from occurring for all anticipated transients. This is P

accomplished by proper setting-of the reactor trip as a function of power level and axial power imbalance. These settings will be given in the Technical Specifications.

3.3 Steady State and Loss-of-Flow Transient The effect of fuel densification on the departure from nucleate boiling ratio (DNBR) during steady state operation was analyzed by.

both the applicant and the staff. The staff's independent calculations are described in Reference 6.

The results show that the steady state minimum DNBR decreases due to an increase in the surface heat flux resulting from fuel-densification. To assess the amount of reduction in DNBR margin, the applicant reanalyzed the steady state operating and design' overpower conditions with an assumed axial power shape that peaked nea the core outlet rather than with the symmetrical reference design power shape described in the FSAR. The outlet shape, though.

not achievable in operation, produces the largest possible DNBR penalty from fuel densification, because the point of =4n4="= DNBR is shif ted

_ toward the top of. the hot fuel: rod where the densification induced power spike is the largest. The application of this large_ power spike at the point of minimum DNBR-produces the greatest degradation

'n DNBR. Using this' outlet axial power peak the applicant computed i

a 4.4% reduction.in _ DNBR from the 1.55 value reported in the FSAR

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. without the effects of densification.. The applicant has proposed-e

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more stringent control rod positions, offset and overpower limits t'o l

compensate for the loss in DNBR margin. This is acceptable to the

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l' staff and will be included in Technical Specifications.

.B&W also reanalyzed the loss of flow trans'ient that would result from a loss'of electrical power to the. reactor coolant-pumps taking-into account the effects of fuel densification. The results show that the minimum DNBR'during tha transient decreased, due to local flux increases caused by fuel densification. The previously_ cal-l culated minimum DNBR during the transient was 1.60 whereas with the densification the minimum DNBR is calcula'ted to be about l'.53.

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The densification effects that_could aggravate the consequences

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j of the loss-of-flow transient are the increase in the steady state i

fuel temperature (stored energy),~ increase in heat flux, and a decrease l

in gap conductance. The increase in fuel temperature provides more l-

-stored heat'in the fuel which,must be removed during the transient; the higher ten flux provides greater initial enthalpy in the coolant channel. The decrease 'in gap conductance delays the re: oval of heat from the fuel resulting in a higher ratio of heat flux ts channel flow during the transient and thus a lower DNBR.

d 3.4 Other'Trancients l

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The following other transients have been reviewed.to determine q

whether-the effects _of densification have resulted in significant changes in their consequences:

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i-3-6 Control Rod Withdrawal Incident Moderator Dilution Incident Control Rod Drop Incident Startup of an Inactive Reactor Coolant Loop

' Loss of Electrical Power-l In theLapplicant's FSAR.these transients were calculated to result.

in a DNBR-in excess of l'.3, or their consequences were shown to be l

limited to acceptable values by limits to be set forth in the Technical

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Specifications.. The staff has reviewed these transients taking into l

account the effects of fuel densification and egrees with the applicant -

l that they would not result in.a reduction of the core therral cargin, I

i.e., a DNBR less than 1.3.

3.5 Conclusicas The effects of fuel densification on steady state and transient 1

operation have been evaluated by the applicant and reviewed-by the staff.

Tae effect on steady state operation, moctly due to local increases in thereal neutron flux and heat generation, is to require more restric-

-tive limits on. control. rod positions and offset limits in the Technical

Specifications for Oconee Unit-3.

In' order to prevent fuel melting the maximum allowable linear heat generation rate has been reduced from 22.2.Kw/ft to 20.15 Kw/ft. The. overpower trip limit has been T

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3-7 reduced fron~114 percent to 112 percent such that.a DNBR greater than 1.3 is maintained for steady' state and during transient conditions.

The staff-concluded on the basis of its review that the potential e'ffects of~ fuel densification on the steady state and postulated-transient operation have been evaluated in an appropriate =anner and are acceptable for.the perivd of operation proposed.

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4-1 4.0 ACCIDC;T A'!ALYSES-4.1 General.

Analyses-of the consequences of verious postulated accidents were presented. in' the FSAR for the Oconee rnit 3.

The accidents evaluated were:

(1) Locked Rotor (2) Loss-of-Coolant (LOCA)

(3) _ Control Rod Ejection (4) Steam Line Rupture (5) Steam Generator Tube Rupture (6) Fuel Handling (7) Waste Gas Tank Rupture Since fuel densification will effect the consequences of the first i

four postulated accidents they have been reanalyzed by the applicant and reevaluated by the staff. Results of the first three accidents (locked rotor, loss-of-coolant, and control rod ejection) are presented in separate parts of this section. The steam generator tube rupture, waste gas tank rupture, fuel handling and steam line rupture accidents

.are discussed below.

- Changes in -the fuel pellet geometry can cause the stored energy ia the fuel pellet to ' increase by the mechanisms discussed in Section 3.0 of this report. Potential increases in local power due to the formation of. axial gaps are discussed in Section 3.1.

Both of these effects are accounted for-in the evaluation of accidents.

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4-2 The radiological consequences of accidents were independently calculated by the ' staff. ' The - results of the staff's calculation of the radiological consequences of accidents were presented in the Oconee Unit 2/3 Safety Evaluation report dated July 6, 1973. The radiological consequences would not increase as a result of fuel densification, a',though the transient parfor=ance of-the fuel rods can change as a result of fuel densification. It is the latter factor that is dis-cussed in the following sections.

The staff evaluation of. the radiological consequences of a vaste gas decay tank failure was based on an assumed quantity of gas'in the tank limited by the Technical Specification. For the steam generator-tube' rupture accident, the assu=ed quantity of reactor coolant activity l

I is consistent with the Technical Specifiestion limits on maximum per-

.mitted reactor coolant system activity.

Fuel densification will not affect'the consequences of these. accidents.

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t The postulated refueling accident assumes the dropping of a fuel i-l assembly in the spent fuel pool'or transfer canal. The fuel rods are i

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-assumed to be approximately ambient temperature during the postulated accident. : Therefore, the direct a'#ects of fuel densification will not -

affect the consequences of this postulated accident. The potential for mechanical failure of a flattened rod might be different from I

that of-a normal ~ rod; however, since the staff evaluation has been l

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fuel' cycle (Sectics 2.0)', this potencial change in fuel rod character--

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.istics was not considered. Furthermore, a'll ofithe rods in the t-dropped; assembly.are assumed to fail.

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- The steam line' break accident was analyzed by the ~ applicant in the FSAR without theJeffects'o'f fuel densification. That. analysis-l showed that the worst consequences from this accident would result i

I at the"end of life (E0L) of the core. Since the DNBR margin is higher at the EOL,(6) including the effects of fuel densification, the staff does not expect that the thermal limits will be more I

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. severe than those presented in the FSAR.

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. Lock'ed Rotor Accident The reactor coolant system for Oconee. Unit 3 consists of two loops; each _ return from. the steam-generator - to the reactor. consists of : two cold: legs,[.e..atotalof'fourreactorcoolantpumpsareused.

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. Locked roter accidents are-characteristically less severe for 4-

- pump plants than for 3 or' 2 pump ' plants.

l The analysis of the-locked rotor accident was originally presented in Section'14:of the FSAR. The transient behavior was en-Irzad by postulating an' instantaneous seizure of'one reactor pump rotor.

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- The reactor-flow would decrease rapidly and.a reactor trip would occur

asLa result of'a'high. power-to-flow signal. The core flow would reduce 1

to about:three fourths its normal full-flow value within two seconds.

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4-4 The 'te=perature,of the' reactor coolant would increase, causing fluid expansion with a resultant pressure transient which would reach a peak of approximately 15 psi above nominal. The applicant cocputed a maxi-0 mum cladding temperature of 1390 F at about 4.0 seconds for this accident.

The staff perfor=ed independent calculations for this postulated accident using Oconee Unit 1 parameters and, confirmation of these calculations are discussed in Reference 6.

4.3 LOCA Analysis The B&W evaluation model described in the AEC Interim Acceptance

. Criteria and Amendments for Emergency Core Cooling Systems was used by the applicant to evaluate the loss-of-coolant accident (LOCA) for Oconee Unit 3.

The analysis was performed with the B&W code CRAFI for the blowdown period and the THETA code for the fuel rod heat up.

The applicant's LOCA analysis without the assumption of fuel densification

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2 is reported in the Oconee FSAR based on the 8.55 f t split break in the cold leg at the pump discharge as the lbsiting break size and location.(8)_

1 During the blowdown period the gap conductance, reduced due to fttel densification according to the staff requirements, could cause 1

i the. core-average fuel pellet temperature to increase, but CRAFT cal-culations show that the temperature experiences only a very small change.

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Since in the initial anal rsis an average core te=perature was used that is higher than the average core te=perature resulting from the decreased gap conducu.nce, the applicant concludes and we agree that the limiting break size and locations do not change due to' fuel densification.

The effects of fuel densification on the reflood calculations is small, since the gap conductance is cuch larger than the film coef-ficient (cladding surface to coolant) durint reflood. The film coefficient is thus limiting with regard to heat transfer and cladding temperature.

The applicant performed ECCS calculations to determine the axially

-l dependent Kw/ft limits for axial peaks located at various elevations in the core from 4 to 10 feet from the core inlet. Operating restrictions will be imposed by the Technical Specifications to limit the actual peak linear heat rate to less than the. axially dependent LOCA Kw/ft limits. The operating restrictions take into consideration fuel depletion, control rod position,' axial power imbalance, transient xenon and quadrant j

power tilt. Administrative controls will be provided to ensure that the LOCA Kw/ft limits are not exceeded during plant operation.

l 4.'4 Rod Ejection Accident

. The control rod ejection transient has been reanalyzed (4,5) by

-the applicant to account for changes in the fuel due to densification.

The significant effects of fuel densification are an increase in the initial maximum fuel temperature and a slight increase in average heat. flux due to shrinkage of the pellet stack length. In

. addition, spikes in the neutron power can occur to gaps in the

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4-6 Calculations (6) have verified that no changes in the basic fuel.

- kinetic response of the core occur due to the small changes in fuel

. geometry and heat transfer characteristics.

The results of the rod ejection accident at BOL and at EOL with-out' consideration of'densification effects have been previously presented in the'Oconee FSAR. The staff consultants at Brookhaven National Laborgtory (BNL) have perfor=ed independent check calcula-tions using appropriate input data and their own computer codes and have confirmed that the results of a rod ejection transient are less severe at EOL than at BOL. Therefore, all calculations by the applicant considering densification effects were done for BOL conditions.

For the full power transient, the control rod reactivity worths available for the assumed ejected rod would be expected to decrease because of the more restrictive insertion limits cn the control bank.

However, this was not included in the_ reevaluation, thereby adding-additional conservatism to the calculations. The maximum Technical-Specification rod worth of 0.65% delta k/k was used.for the BOL calculationr at full power.

The staff review of:ths iv.itial fuel temperature for the BOL full power case indicated that a reasonable temperature-was used for the

' assumed-conditions, consistent with that used in the LOCA analysis.

The neutron poker-spike effect was. included in the reanalysis.-

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t 4.-7 The reexamination of the rod ejection transient considering' the effects of densification has resulted in a peak peller average enthcipy well below the staff's criterion of 280 cal /gm. The maximum center-line. fuel' temperature reached is well below the assu=ed melting point of 5080 F, and the maximum clad temperature during the transient is 15600F. The total number of fuel pins calculated to be in DNB is 28%.(14) h Mfmbdmw g % gg g reasonably conservative consideration has been given to the effects of fuel densification and that the results are acceptable for this accident.

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5-1 5.0 SUmfARY AND CONCLUSIONS

.The effects of fuel densification have been considered in analyses of normal operation, operation during transient conditions, and postu-11ated accident conditions. On the. basis of.the staff review of the applicant's calculations, and independent calculations performed by the staff and its consultants, the staff concluded that for the period of operation proposed, namely the first fuel cycle:

(1) The effects of densification during steady state and transient operation of the Oconee Unit 3 reactor will not cause the limits on DNBR, cladding strain, and centerline temperatures, to become less conservative than values previously established in the FSAR.

'(2) The effects of densification were included in the-calculation of fuel rod behavior during postulated loss-of-coolant accidents.

The LOCA analysis is acceptable and cotaplies with the June 1971 Interim Acceptance Criteria.

(3) The applicant's omission of the creep down effect, which tends to increase gap conductance with life time, is acceptable.

(4) The' Technical Specifications will limit the fuel residence time to 10,944 effective full power hours of power operation to assure no cladding collapse.

-(5) The applicant has adopted the staff reco::nnendations for calcu-

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lating gap conductance and fuel temperatures (Section 3.2) as-

-they are used in. steady. state, transient and accident conditions.

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4 (6) _ Operating restrictions as necessary to assure compliance with items-(1) through (4) above will be incorporated into the T

Technical Specification.

On the basis'. of the above su=ary, the staff concludes that the

--applicant-is_in compliance with the staff densification report (1) and that Oconee Unit 3 reactor can be operated at_ power levels up to 100%

of rated power with no undue risk to the health and safety of the public.

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L 6.' O REFERENCES

- 1.

Technical Report on Fuel Densification of Light Water Reactor j

Fuels," Regulatory Staff, U.S. Atomic Energy Commission, November 14,

~1972.

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" Fuel Densification Report," BAW-10054 Topical Report (Proprietary),-

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January 1973 (Nonproprietary Information in BAW-10055).

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"Oconee 1 Fuel Densification Report," BAW-1387 (Proprietary),

January 1973 (donproprietary Information in BAW-1388).

'4.

" Fuel Densification Report," BAW-10054 - Rev. 1 Topical Report (P sprietary), April 1973.

i 5.

"Oconee 1 Fuel Densification Report," BAW-1387 - Rev. 1 (Proprietary) l April'1973.

6. '" Technical Report on Densification of Babcock. & Wilcox Reactor Fuels" by the Regulatory ataff, U.S. Atomic Energy Commission, l

July 6, 1973.

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LetterfromR.C.beYoungtoR. Edwards, Babcock &Wilcoxdated L

April 23,.1973, with copy to Duke Power Company.

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"Multinode Analysis of B&W's' 2568 MWT Nuclear Plants During a l'

L Loss-of-Coolant Accident," BAW-10034, Rev. 1, May 1972.

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9. ' Letter from Duke Power Company to A. Giambusso, dated May. 14, 1973.

. 10. "TAFY - Fuel Pin Ttgerature and Gas Pressure Analysis," BAW-10044,

.l Topical Report, April-1972.

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11. Oconee 2 Fuel. Densification Report," BAW-1395 (Proprietary)

' June 1973.-

- 12. - Peaking Factors:in Pressurized Water Reactors with Fuel Densification"~

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BNL Interim Report, Dece=ber 1972.

13. '.' Peaking Due to Densification in the Ifaine Yankee Reactor"; BNL Interim Report,>tarch 1973.

. 14. '"Oconee 3 Fuel Densifica' tion Report," BAW-1399-(Proprietary)'-

-December 14, 1973..

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