ML20074A815

Forwards Endorsement 44 to Nelia Policy NF-236 & Endorsement 34 to Maelu Policy MF-92

ML20074A815
Person / Time
Site:	Davis Besse
Issue date:	05/02/1983
From:	Golden M MARSH & MCLENNAN, INC.
To:	Saltzman J NRC OFFICE OF STATE PROGRAMS (OSP)
References
NUDOCS 8305160105
Download: ML20074A815 (3)

Text

.

Marsh &

. . m Mclennan Marsh & McLennan, incorporated 1221 Avenue of the Americas New York, New York 10020 Telephone 212 997-2000 May 2, 1983 Mr. Jerome Saltzman Assistant Director of State and Licensee Relations Office of State Programs U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Toledo Edison Company ANI/MAELU Policies NF-236, MF-92

Dear Jerry:

On behalf of the Toledo Edison Company, I have enclosed eight certified copies each of Endorsement 44 to NF-236 and Endorse-ment 34 to MF-92.

Please contact me if there are any questions.

Very truly yours, /

w f/y Michael P. G61 den Nuclear Consultant MPG:aj Enc.

cc: R. Ertle 8305160105 830502 PDR J ADOCK 05000346 pyg

,, . Nuclear Energy Liability insurance NUCLEAR ENERGY LIABILITY INSURANCE ASSOCIATION ADVANCE PREMIUM AND STANDARD PREMIUM ENDORSEMENT CALENDAR YEAR 1982 It is agreed that Items la. and Ib. of Endorsement No. 42 are amended to read:

Ia. ADVANCE PREMIUM: It is agreed that the Advance Premium due the companies for the period designated above is: $289,323.00 .

Ib. STANDARD PREMIUM AND RESERVE PREMIUM: In the absence of a change in the Advance Premium indicated above, it is agreed that, subject to the provisions of the Industry Credit Rating Plan, the Standard Premium is said Advance Premium and the Reserve Premium is: $ 217,560. 48 .

Return Premium: $ 2,893.24 .

This is to cortify that this is a true copy of the original Endorsement having the endorcement number and being made part of the Nuclear Energy Liability Policy (Facility Form) as des-ignated hereon. No Insurance is affcrded hereunder.

John L Quattrucchi, Vim President Liability Underwriting l American Nudear Insurers Effective Date of January 1, 1982 NF-236 a pa o cy No 12:01 A.M. Standard Time issued to The Toledo Edison Company Date of issue Apri1 21, 1983 For the su icnbing co panies By )

/3 General Manager Endorsement No 44 Countersigned by NE-36 L

NUCLEAR ENERGY LIABILITY INSURANCE MUTUAL ATOMIC ENERGY LIABILITY UNDERWRITERS

1. AMENINENT T AWANCE PRDiIUM DJDORSD4Dir

2. SPANDARD PREMIUM AND RESERVE PREMIUM D700RSDETT

3. REIURN PRDiIUM WE

1. Advance Pranitzn It is agreed that the Amended Advance.Franian due the cxr:panies for the calendar year 1982 is $83,997.00 .

2. Standard Prenian and Reserve Preniun Subject to the provisions of the Industry Credit Rating Plan, it is agreed that the Standard Pruniun and Reserve Preniun for the calendar year designated above are:

Standard Praniun $83,997.00 Reserve Praniun $63,162.72

3. Retuzn Premian $839.97 .

Effective Date of 'Ib fonn a part this Endorsement January 1, 1982 of Policy No. MF-92 Issued to The Toledo Edison Company Date cf Issue April 21, 1983 i

i l

l For the Subscribing Carpanies t

I i WILE A'ITMIC DJEPGY LIABILITY UNDEli By -

,y/ A Q

Endorsement No. 34 Countersigned by i .

Authorized Pepresentative This is to certify that this is a true copy of the original Endorsement having the endorsement number and being made part of the Nuclear Energy Liability Policy (Facility Form) as des ME-36 Ignatnd hereon. No Insurance is afforded hereunder.

l $.

l John L. Quattrocchi. Vice President Liabihty Underwriting American Nuclur Insunra

A Attachment B to AEP:NRC:0745C Safety Evaluation of Reload

1.0 INTRODUCTION

D. C. Cook Unit 1 is operating with an all Exxon Nuclear Company (ENC) fueled core during Cycle 7. For subsequent cycles, it is planned '.o refuel Unit 1 with 15x15 optimized fuel assembly (OFA) regions supplied by the Westinghouse Electric Corporation (W). As a result, future core loadings would range from approximately a 40% OFA and 60% ENC fueled core to eventually an all 0FA fueled core. The W 15x15 0FA fuel

! cesign is similar to the W 15x15 LOPAR (low parasitic) fuel which has had substartial coerating perfomance in a nt.mber of nuclear plants.

The major difference introcuced by the f 15x15 0FA design is the use of five intercediate Zircaloy grids replacing five intermediate Inconel grids for the LCPAR fuel. The 15x15 Zircaloy grid design is similar to the W 17x17 OPA grid design. The W 17x17 M A cetign has been generi-cally approved by the NRC via their raview of the W 17x17 GFA Reference Core Reoort.0) Operating experience has been obtained for six demon-stration 17x17 0FAs which contain Zir'czioy intermed! ate grids.IE) Two i

assemblies have satisfactorily completed three cycles of irradiation to about 28,0^)0 MWD /MTU burnup, two have comp;eted tw$ cycles to about 19,400 MWD /MTU, and tw: have completed one cycle in excess of 9,000 MWD /MTU. The dsmoratration 0FAs have been examined anc pre, vide reason to expect good performance frem the 15x15 0FA desi p .

l This report summarizes the results of the W analyses which justify the

transition from an all ENC core, through a mixed 0FA/ ENC fueled core to an all 0FA core. Although it is planned to operate D. C. Cook Unit 1 Cycle 8 at the current licensed maximum power level of 3250 MWt, the core evaluations / analyses summarized in this report have been performed

, at a reactor power level of 3411 MWt, with the exception of the large break LOCA which was analyzed at 3250 MWt. This conservative design basis provides early identification of those safety / accident analysis limits for a potential uprating.

l 1 .

s 4

\

All analyses were performed utilizing W standard methods, which are described in the W Reload Safety Evaluation Methodology Topical.(3)

The approved Westinghouse Improved Thermal Design Procedure (ITDP) is used in the DNB analyses of both W and ENC fuel. The W WRB-1 correlation is used in the OFA DNB analyses. Both the ITDP and WRB-1 correlation were p'reviously used to license D. C. Cook Unit 2 operation. The ENC fuel is analyzed using the W-3 DNB correlation. Other features being introduced with the Cycle 8 reload include the Westinghouse Wet Annular Burnable Absorber (WABA) rods and a revision to the Westinghouse fuel thermal safe-ty model (PAD Code) used in the safety analyses. Westinghouse has sub-mitted topical reports ( ,5) on these sLbjects and is supporting the NRC's generic review, in order to obtain approval well befora the planned Cycle 8 startup.

S-O l

l l

2

+

2.0

SUMMARY

AND CONCLUSIONS The Westinghouse Reload Safety Evaluation Methodology (3) was used to evaluate the transition from ENC fuel to W 15x15 0FA fuel for D. C.

Cook Unit 1. Parameters were chosen to maximize the applicability of the transition evaluations for each reload cycle and to facilitate the safety evaluation of future reload Cores. Iransition Core effects were considered in the mechanical, thermal and hydraulic, nuclear, and acci-dent evaluations described in Chaoter 18 of Reference 1. The summary of these evaluations for the D. C. Cook Unit 1 transition to an all W 15x15 0FA core is given in the following sections of this submittal.

The transition design and safety evaluations are based on the following maximum power conditions: 3411 MWt reactor power and 577.1"F vessel average temperature.

I The results of evaluations / analyses and tests discussed in tnis report lead to the following conclusions:

1. The Westinghouse OFAs are me:hanically and hydraulically compatible with the ENC fuel assemblies, ' control rods, and reactor internals interfaces.

l l 2. Changes in the nuclear characteristics due to the transition from ENC to W 15x15 0FA fuel will be within the normal variations from l ,

cycle-to-cycle due to fuel management effects. W 15x15 0FA fuel up to and including a 4.00% nominal enrichment can be stored in the fresh and spent fuel areas.

3. Demonstration experience with W 17x17 0FAs containing Zircaloy grids provides reason to expect satisfactory operation from 15x15 0FA Zircaloy grids.

I I

e

4. The WABA rod, as described in its generic topical (4) , is compa-tible with the W 15x15 0FA and satisfies all performance require-ments for its design life.

5. The propssed Technical Specification changes presented in Attach-ment A are applicable to cores containing any combination of W 15x15 0FA and ENC fuel.

6. All design criteria far the W 15x15 0FA fuel are satisfied.

7. A reference is established upon which to base future cycle safety ,

evaluations for V 0FA reload fuel.

i *g a

< g L

d 4

3.0 MECHANICAL EVALUATION The mechanical design requirements and criteria for the 17x17 0FA design are described in Reference 1, which was approved by the NRC. The 15x15 0FA design meets these same basic requirements and criteria.

ENC, in estabishing their assembly design, demonstrated their fuel's compatibility with the W LOPAR design which was the initial D C. Cook Unit I fuel. W has demonstratec compatibility of its 15x1E OFA design with its LOPAR design. Compatibility of the OFA and ENC fuel is thereby demonstrated.

Figure 1 and' Table 1 present a comparison of the W 15x15 0FA and ENC fuel assemblies. The W and ENC fuel rods have similar length and clad OD dimensions. The W 15x15 0FA rods have the same design as the LOPAR W 15x15 fuel rods which have exhibited good in-core performance in many operating reactoes.

The top and bottom Inconel grids of the CFA are the same as the Inconel grids of a V LOPAR fuel assembly. The five intermediate OFA Zircaloy-4 grids have thicker and wider straps than the OFA Inconel grids (See Figure 1) in order to closely duplicate the Inconel grid strength. The ENC assembly grids are bimetallic, consisting of Zircaloy-4 straps with Inconel grid springs. . Both the OFA Zircaloy and ENC bimetallic grids have grid heights of 2.25 inches. Elevation of the grids was estab-lished to ensure satisfactory axial alignment during operation.

Due to thicker Zircaloy grid straps and a resulting reduced cell size, the OFA guide thimble tube ID (above dashpot) has a 12 mil reduction l

compared to the ENC thimble tube ID of 0.511 inches. Below the dashpot, the OFA and ENC fuel thimble tubes have the same dimensions. The OFA guide tube thimble ID provides sufficient nominal diametral clearance for control rods as well as source rods, burnable absorber Vodi, and 0FA thimble plugs. Due to reduced 0FA diametral clearance, the control rod 5

^

l scram time to the dashpot is increased from the current 1.8 seconds to 2.4 seconds. This increase in rod drop time was determined from conservative analytical calculations. The 2.4 second scram time is used in all the accident reanalyses'.

The OFA design has minor differences in the overall height of the top .

and bottom nozzles, the adapter plate flow-slot configuration and hold-down lea? springs as compared to the ENC fuel assembly design. These minor differences base no adverse impact on the interaction of W 15x15 0FA and ENC assemblies during fuel handling operations or reactor opera-tions. The W 15x15 0FA design uses a 3-leaf holddown spring design compared to the 2-leaf springs in the ENC assembly. The W OFA 3-leaf spring has been previously used in 15x15 LOPAR assemblies, as well as on the 17x17 0FA demonstration assemblies. The 3-leaf spring provides additional holddown force margin compared to the 2-leaf spring. The OFA bottom nozzle has similar design features and dimensions compared to the ENC nozzle. The OFA bottom nozzle design has a reconstitutable feature,

! as shown in Figure 2, which allows it to be easily removed. A locking cup is used to lock the thimble screw of a guide thimble tube in place, instead of the lockwire as used for the standard W LOPAR nozzle design. The reconstitutable nozzle design facilitates remote removal of the bottom nozzle and relocking of thimble screws as the bottom nozzle 1

is reattached.

As stated in the 17x17 0FA Reference Core Report , for a given l burnup, the magnitude of rod bow Yor the W OFA is conservatively assumed to be the same as that of a W LOPAR fuel assembly. The most

~

probable causes of significant rod bow are rod grid and pellet-clad interaction forces and wall thickness variation. Since the OFA fuel rods are the same as the W LOPAR fuel rods, there will be no difference l in predicted bow due to rod' considerations. The OFA design will have reduced grid forces due to the Zircaloy grid springs. Therefore, this component is predicted to decrease OFA rod bow compared to LOPAR fuel.

6

The wear of fuel rod cladding is dependent on both the support provided

~

by the grids and the flow environment to which it is subjected. OFA and ENC assembly flow test results were evaluated. ENC hydraulic test results show the cross flow between ENC and W 15x15 LOPAR assemblies is very similar to that obtained during W flow tests on side-by-side W 15x15 0FA and W 15x15 LOPAR assemolles. These tests showed only a  :

small cross flow between assemblies and no significant fuel rod wear due to rod vibration. Extrapolation of the results from flow tests involving 0FA and LOPAR assemblies shows that fuel rod wear would be less than ten (10) percent of the cladding thickness for at least 48 months of reactor operation. This assures that clad wear will not impair fuel rod integrity.

The above conclusions on 0FA rod wear and integrity have also been sup-ported by analytical results. The analysis accounted for rod vibrations caused by both axial and cross flows, and the effect of potential fuel rod to grid gaps.

1 i

.-e=

7 l

4.0 NUCLEAR EVALUATION The nuclear design of cores with y 0FA and ENC fuel is accomplished by using the standard calculational methods as described in the W Reload Safety Evaluation Methodology (3) . The dimensional and material dif-ferences between the y and ENC assemblies are small so that the y computer codes and methods are also valid for the ENC fuel. Dimensions and composition for each of the two fuel designs were used to establish the models. The burnup distribution of the ENC fuel assemblies remaining in Cycle 8 has been obtained by depleting the loading patterns from earlier cycles using two dimensional and three dimensionai models of the applicable cores.

Changes in the nuclear characteristics during the transition cycles from an ENC fueled core to a y 15x15 0FA core will be primarily due to fuel management considerations (number of feed assemblies, feed enrichment, cycle burnup, etc.) and not due to the differences in fuel assembly design. Each reload core design will be evaluated to assure that design and safety limits for the OFA and ENC fuel are satisfied according to the W reload safety evaluation methodology. For the evaluation of the worst-case F (Z) envelope, axial power shapes are n

synthesized with the limiting Fxy values chosen over three overlapping burnup windows during the cycle. The design and safety limits will be documented in each cycle specific reload safety evaluation report which serves as the basis for any significant changes requiring NRC review.

In order to accommodate potential increases in future feed enrichments, a criticality analysis of the fuel storage areas was performed for nomi-nal enrichments up to and including 4.00 Wt.% U235 in W 15x15 0FA fuel. These analyses confirm that all current safety criteria applicable to fuel storage are satisfied (6) ,

8

0 5.0 THERMAL AND HYDRAULIC EVALUATION Results of hydraulic compatibility tests performed by the Exxon Nuclear Company for the ENC and W 15x15 LOPAR assemblies were compared to hydraulic test data for the W 15x15 LOPAR and 0FA assemblies. The data show that the W 15x15 0FA fuel assemblies are hydraulically compatible with the ENC fuel assemblies. Pressure drop data were obtained over a range of fluid temperatures and flow rates. Pressure drops values were then extrapolated to core operating conditions. At typical reacter conditions, the ENC fuel assembly has a pressure drop within 0.7 percent of the W 15x15 0FA pressure drop.

The thermal hydraulic design of this core is conservatively analyzed at 3411 MWt core power with a 577.1*F vessel average temperature, even th Ngh the Cycle 8 core will continue to be limited to its current rated parameters of 3250 MWt core power and a 567.8*F vessel average tempera-ture. The analyses employed the Improved Thermal Design Procedure (7)

(ITDP) and the THINC IV(8,9) computer code. The WRB-1(10) DNB cor-relation was used in the W 15x15 0FA analyses, whereas the W-3 correla-tion was used to analyze the ENC fuel. The thermal hydraulic design criteria remain the same as those presented in the D. C. Cook Unit 1 Updated FSAR(11) . All design criteria are satisifed.

The design method employed to meet the DNB design basis is the ITDP(7) . Uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are consider'ed sta-tistically, such that there is at least a 95 percent probability that the minimum DNBR will be greater than or equal to the limit DNBR for the peak power rod. Plant parameter uncertainties are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined with the DNBR limit, establishes a design DNBR value which must be met in plant safety analyses. Since the parameter uncertainties are considered in deter-mining the design DNBR value, $ne plant safety analyses are performed using values of input parameters without uncertainties. In addition, 9

b ,o O

the limit DNBR values are increased to values designated as the safety analysis limit DNBR's. The plant allowance available between the safety analysis limit DNBR values and the design limit DNBR values is not required to meet the design basis.

In this application, the WRB-1 DNB correlation (10) is employed in the thermal hydraulic design of the W 15x15 0FA fuel. Due to an improve-ment in the accuracy of the critical heat flux prediction with the WRB-1 correlation compared to previous DNB correlations, a correlation limit DNBR of 1.17 is applicable. The W-3 DNBR correlation (12,13) was used in the design of the ENC fuel assembly. A W-3 correlation limit DNBR of 1.30 is applicable.

The table below indicates the relationships between the correlation limit DNBR, design limit DNBR, and the safety analysis limit DNBR values used for this design.

W 15x15 0FA ENC 15x15 Typical Thimble Typical Thimble

___ Correlation Limit 1.17 1.17 1.30 1.30 Design Limit 1.32 1.31 1.58 1.50 Safety Analysis Limit 1.69 1.69 1.58 1.50

'The margin to.the safety analysis DNBR limit is more than sufficient.to cover the maximum 12.5 percent rod bow penalty at full flow Conditions (14) and a 5 percent transition core penalty, both applied to the OFA only. An additional rod bow penalty of 2.4% DNBR at loss of flow conditions (14) is covered explicitly in the loss of flow analysis for the W 15x15 0FA. The 5 percent transition penalty was determined by analyzing W 15x15 0FA and ENC l

I i

l l

10

e assembly loading patterns at various core conditions in the same manner as the W 17x17 0FA/LOPAR fuel analysis which was reviewed and approved by the NRC(15), The 5 percent transition penalty for OFA is due to the higher 0FA mixing vane loss coefficient compared t6 that of the ENC fuel. This results in localized flow redistribution from the OFA to the ENC assembly near mixing vane grid positions. When the full transition is complete (all ENC assemblies removed from core), the transition core penalty will no longer apply to 0FA assemblies. .

The ENC fuel assembly would be expected to have less gap closure than the W 15x15 0FA, due to 'the ENC fuel's thicker cladding, as shown in Reference 16. Data obtained by other investigations (17,18) show that gap closures up to 55% have no measurable effect on DNB. Therefore, no resultant' rod bow DNBR penalty is required for ENC 15x15 fuel.

P I

o W

r b

1 -

1 11 .

o 6.0 ACCIDENT ANALYSES AND EVALUATION 6.1 NON-LOCA ACCIDENT ANALYSES AND EVALUATION The effects of the transition from the resident ENC fuel to W OFA on the non-LOCA accident analyses have been addressed. The standard West-inghouse reload methodology described in Reference 3 was used. All of the non-LOCA accidents

• in the D. C. Cook FSAR were reanalyzed to incl.ude three major design changes:

1. The analyses were performed at a conservative reactor power level of 3411 MWt. Tnis affects all of the transients that are limiting at full power.

2. The ITDP was used with both the WRB-1 and WRB-3 DNB correlations.

This impacts all of the DNE limited accidents. A conservat1ve set -

of core thermal safety limits overtemperature delta T and overpower delta T setpoints were generated that are applicable for both the transition ard ccaplete OFA cores. These limits are valid fer reactor power levels up to and including 3411 MWt.

3. The control rod scram time to the dashpot is increased from 1.8 seconds to 2.4 seconds. This increased drop time primarily affects j the fast reactivity transients but was used in all of the analyses

! requiring this parameter.

Also included in the analyses were fuel temperatures based on the revised PAD code. A +5 pcm/ degree F moderator temperature coefficient (MTC) existing at full power was conservatively used for heatup events.

This is conservative since the Technical Specifications require a non positive MTC at or above seventy (70) percent power.

• With exception of startup on an inactive loop. This transient cannot occur above 10% rated thermal power and thus was not reanalyzed.

~~-s "Nr

V U

~

4 The acceptance criterion used in the non-LOCA safety analyses is independent of fuel. vendor. Thus, the results of the FSAR Chapter 14 accident reanalysis and evaluation, which are contained in Attachment C, show that the transition to 0FAs can be accommodated with margin to the applicable FSAR safety limits for power levels up to and including 3411 MWt.

6.2 LARGE BREAK LOCA (@ 3250 MWt)

Description of Analysis Assumptions for W 15x15 0FA Fuel, Including Transition Impact The large break loss-of-coolant accident (LOCA) analysis for D. C. Ceok

'Jnit 1, applicable to a full W 15 x 15 0FA core, was analyzed to deve-lop W 15 x 15 0FA fuel specific peaking factor limits. This ane. lysis is consistent with the methodology employed in Reierence 1. Thy cur-rently approved 1981 large bnak. ECC$ evaluation modelO9) was utt-lized for a spectrum of. cold lag breaks. The revised PAD fuel thermal safety model(5) generated the initial fuel rod conditions. The [' C.

. Cook Unit 1 analysis w6s performed for an assumed ' steam generator tube plugging level of five (5) percent, and was analyzed Ne both minimum and maximum safeguards (safety injection flows) assumptions, in accor-dance with Reference 20. A revised FSAR chapter 14.3.1.1, g!ven in Attachment D, centains a full description of the analysis and assump-tions utilized for the W OFA ECCS LOCA analysis. The ENC fuel ECCS l

analysis contained in FSAR section 14.3.1.2 remains unchanged.

When assessing the LOCA impact of transition cores, it must be deter-mined'whether the transition core can have a greater calculated peak clad temperature (PCT) than either a complete core of the reference fuel 13

design or a complete core of the new fuel design. For a given peaking factor, the only mechanism available to cause a transition core to have a greater calculated PCT than a full core of either fuel is the possi-bility of flow redistribution due to fuel assembly hydraulic resistance mismatch.

For the ENC and W 15x15 0FA designs, this difference in fuel assembly resistance (K/A2 ), is less than one percent. The different flow J

resistances for the two assembly designs in. pact two portions of the LOCA analysis model. One is the reactor' coolant system (RCS) blowdown por-tion of the transient, analyzed with the SATAN VI computer code, where the higher resistance W OFA assembly has less cooling flow than the ENC assembly. While the SATAN VI computer cede models the cross flow between the average core flow channel (N-1 fuel assemblies) ano a hot assemb'y flow channel (one fuel assemoly), experience has shown tnat the SATAN VI results are not significantly affected by small differences in ,

the hydraulic resistance (110%) between these two channels. Since small resistance mismatches in the core are insicnificant when compared to the total system resistance, and since the totrl core rasistance is 3 uniformly distributed in the SATAN VI cede, the efftet on the large break LOCA blowdown transient of modeling hydraulic resistance mismatch can be reglected. Therefore, it is not r;ecessary or meaningful to per-form a new SATAN VI cnalysis for this transition core configuration i because the hydraulic resistance mismatch is much less than 1 10 per-cent.

The other portion of the LOCA evaluation model impacted by the hydraulic resistance difference is the core reflood transient. Since the hy-draulic mismatch is so small, only crossflows due to the smaller rod i size and different grid designs need to be evaluated. The maximum re-flood axial flow reduction for the W 15x15 0FA fuel at any location in the core, resulting from crossflows to adjacent ENC assemblies, has been conservatively calculated to be three percent. Analyses have been per-9 l .

14 l-l

o o -

J formed, which demonstrate that a reduction of five (5) percent in reflood axial flowra.te results in a 19 F PCT increase. Therefore, the maximum PCT penalty possible for W 15x15 0FA fuel during the transition period is 12*F. After this transition, the W ECCS anaTysis will apply to a full core without the crossflow penalty.

The resident ENC fuel is shown to have axial flowrates always greater than the nominal design flowrate, for core axial elevations where PCT's can possibly occur. Therefore, the ENC ECCS analysis is not detri-mentally affected by assembly crossflow and remains applicable to the ENC fuel for transition cycles.

The method of analysis, including assumptions and codes used, are I

described in detail in the revised FSAR Chapter 14.3.1.1 provided in

• Attachment D. ,

The results of this analysis, including tabular and plotted results of the break spectrum analyzed, are provided in Attachment D.

._Conclusiens For brecks up to and including the double ended severance of a .*eactor coolant pipe, the emergency core cooling system will meet the acceptance l criteria as presented in 10 CFR 50.46. That is:

1. The calculated peak fuel element clad temperature is below the re-quirement of 2200*F.

2. Th'e amount of fuel element cladding that reacts chemically with water or steam does not exceed one (1) percent of the total amount of Zircaloy in the reactor.

l h

l 15 l -.

O ~

4

3. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The localized cladding oxi-dation limit of seventeen (17) percent is not exceeded during or after quenching.

'4. The core remains amenable to cooling during and after the break.

5. The core temperature is reduced and decay heat is removed for an extended period of time as required by the long-lived radioactivity remaining in the core.

The time sequence of events for all breaks analyzed is shown in Table 14.3.1-6 of the revised FSAR Chapter 14.3.1.1, presented in Attachment D. ,

The large break y 15x15 0FA LOCA analysis for D. C. Coot Unit 1 utili-zing the currently approved 1981 evaluation models rescited in a PCT of 2170'F for the 0.4 CD (discharge coefficien;) LOCA Harimum Safeguards Injection (Max. SI) case at a total peaking factor of 2.00 1

Jhe smell frpact of crossflow for transition core cycles is conser-vatively evaluated to be at most a 12*F effect on the y fuel, whicn is easily accommodated in the margin to 10 CFR 50.46 limits.

The ENC EC,CS analysis is not detrimentally affected by assembly cross-flow; consequently the ENC peaking factor limits remain valid for the ENC fuel during the transition perico.

It can be seen from the results contained in Chapter 14.3.1.1 of the j revised FSAR section that this.cCCS analysis for D. C. Cook Unit I remains in compliance with 10 CFR 50.46 of Appendix K.

l 16

4 -

6.3 SMALL BREAK LOCA (@ 3411 MWt)

Description of Analysis Assumotions for 15x15 0FA Fuel Including Transition Imoact The small break loss-of-coolant accident (LOCA) analysis for D. C. Cook Unit 1, applicable to a full W 15x15 0FA core, was analyzed to develop W 15x15 0FA fuel specific peaking factor limits. This is consistent with the methodology employed in Reference 1. The currently approved October 1975 small break ECCS evaluation model(21) , was utilized for a spectrum of cold leg breaks. The revised PAD fuel thermal safety model(5) , generated the initial fuel rod conditions. Revised FSAR chapter 14.2.2, given in Attachment E, contains a full description of the analysis ard assumptions utilized for the W OFA ECCS LOCA analysis.

When assessics the impact of a LOCA on transition cores it must be detsemined whether the transition core can have a greater calculated peak clad temparatura (PCT) than either a complete core of the reference faal design or a cenplate core of the imoroved fuel design. For a given puaring factor, the only mechar> ism availabla to cause a transition core

~

l to have a greater calculated .CCT than a full core of either fuel is the passibility of flow redistribution due to fuel assembly hydraulic resis-tance mismatch. -

The WFLASH computer code (21) 1s uss.d to nodel the core hydraulics during a small break event. Cely one core flow channel is modelled in WFLASH since the core flowrate during a small break is relatively low and this provides enough time to maintain flow equilibrium between fuel assemblies (i.e. cross flow). Therefore, hydraulic resistance mismatch is not a factor for small break. Thus it is not necessary to perform a small break evaluation for transition cores, and it is sufficient to reference the small break LOCA for the complete core of the W 15x15 0FA design.

The methods of analysis, including assumptions and codes used, are described in detail in the revised FSAR Chapte- 14.3.2 in Attachment E.

The results of this analysis, including tabular and plotted results of the break spectrum analyzed, are provided in Attachment E.

Conclusions The small break optimized fuel LOCA analysis for D. C. Cook Unit 1, utilizing the currently approved 1975 Small ' Break Evaluation model, resulted in a peak clad temperatere of 1630 F for the 4 inch diameter ..

cold leg break. The analysis assumed the vorst small break power shape consistant with a LOCA gF envelope of 2.32 at core midplane elevation and 1.5 at the-top of the core.

Analyses presented in the revised FSAR Chapter 14.3.2 sh:w that the high heaa portion of the ECCS, together with the accut.xlators, provide suf-ficient core flooding to keep the calcuisted peak clad toeperature well h? low the required limits of 10 CFR 50,46. Adequate protection is thurefore afforded by the ECCS in the event of a small break LOCA.

t E

o k

18

s 7.0 TECHNICAL SPECIFICATION CHANGES Based ori the preceeding evaluations, a number of technical specification changes for D. C. Cook Unit 1 are required to support the transition to 0FA. These changes are giv'en in the proposed Technical Specification page changes (see Attachment A of this submittal).

1 l

i 1

i f

l 1

e 19

O

8.0 REFERENCES

1. Letter from R. L. Tedesco (NRC) to T. M. Anderson (Westinghouse),

Safety Evaluation of WCAP-9500, " Reference Core Report - 17x17 Optimized Fuel Assembly," NRC SER letter dated May 22, 1981.

2. Jones, R. G. , Ior.ii, J. A., " Operational Experience with Westing-house Cores (up to December 31, 1981)," WCAP-8183, Rev. 11, May 1982.

3. Bordelon, F. M., et. al., " Westinghouse Reload Scfety Evaluation Methodology," WCAP-9273 (Non-Prop.), March 1978.

4. Rahe, E. P;, (Westinghouse) letter to C. O. Thomas (NRC) Number NS-EPR-2670, October 18, 1982,

Subject:

" Westinghouse Wet Annular Bernable Abscrber Evaluation Report," WCAP-10021, Revision 1 (Pro-prietary).

6. Ac.he, E. P. , Westieghouse Letter to C. O. Thomas of NRC, Letter Nuaber kS-E?R-2671, October 27, 1982,

Subject:

" Westinghouse

__ Revised PAC Ccde Thennal Safety Model," WCAP-8720, Addendum 2 (Fro-priets r/) .

6. Hun".er R. 5. to Denton, H. R.,

Subject:

Fuel Sterage Technical Specif,1ce?. ton Change Requast; AEP:NRC:07458, February 28, 1983.

7. Chelemer, H. , et. al . ,' " Improved Thermal Design Procedure,"

WCAP-8567, July 1975.

8. Chelemer, H., et. al., "THINC IV - An Improved Program for Ther-mal-Hydraulic Analysis of Rod Bundle Cores," WCAP-7956, June 1973.

20

s*

8.0 REFERENCES

(Continued)

9. Hochreiter, L. E., et. al., " Application of THINC IV Program to PWR Design," WCAP-8054, September 1973.

l'O. Motley, F. E. , et. al . , "New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids," WCAP-8762, July 1976.

11. D. C. Cook Unit 1 Updated FSAR, updated thru 1982.

12. Tong, L. S., " Critical Heat Fluxes in Rod Bundles, Two Phase Flow and Heat Transfer in Rod Bundles," Annual Winter Meeting ASME, November 1969, p. 3146.

13. Tong, L. S. , " Boiling Crisis and Criter!3 Heat Flux, . . . ," AEC Of-fi:e of Information Services, TID-25887, 1972.  :

14. Stoltz, J. F. , NRC Letter to T. M. Andersen, Westinghouse, " Staff

__ Review of WCAP-8691," April 5,1979.

15. Letter from C. O. Thomas (NRC) to E. P. Rahe (Westinghouse), Sub-

.iect: Supplemental Acceptance Nunoer 2 for Referencing of Licensing  ;

Tco;ca? Recort WCAP-9500, NRC SER Letter dated January 24, 1983.

16. Letter frem G. F. Owsley (ENC) to T.A. Ippolito (NP.C);

Subject:

XN-75-32, Supplement 1, "Computaticnel Procedure for Evaluating Fuel Rod Bowing," July 17, 1979.

17. - Matkowski, et. a1. , "Effect of Rod Bowing of CHF in PWR Fuel Assem-blies," ASME paper 77-HT-91.

21

4

8.0 REFERENCES

(Continued)

18. Lett'er, J. H. Taylor to S. A. Vargo,

Subject:

Status Report on R&D Programs described in Semi-Annual Topical Report BAW-10097A; Rev. 2, November 13, 1978. -

19. Rahe, E. P., " Westinghouse ECCS Evaluation Model, 1981 Version,"

WCAP-9220-P-A (Proprietary), WCAP-9221-A (Non-Proprietary), Revi-sion 1, 1981.

20. Rahe, E. P. (Westinghouse) letter to R. L. Tedesco (NRC), Letter Number NS-EPR-2538, December 22, 1981.

21. Skwarek, R. J. , Johnson, W. J. and Meyer, P. E. , " Westinghouse Core Cooling System Small Break," October 1975 Model, WCAP-8970-P-A (Pro-prietary) and WCAP-8971-A (Non-Proprietary),1977.

I 9

22

0 TABLE 1 Comoarison of 0FA and ENC Assembly Design 15x15 W 15x15 Optimized Fuel ENC Fuel Parameter Assembly Design Assembly Design Fuel Ars'y. Length, in. 159.765 159.71 Fuel Rod Length, in. 151.85 152.07 Assembly-Envelope, in. 8.426 8.426 Compatible with Core Internals Yes Yes Fuel Rod Pitch, in. 0.563 0.563 Number of Fuel Rods / Ass'y. 203 204 Nember of Guide Thimbler/.6ss'y. 20 20 NLmbar of InstrumentEtica Tu' ce/A%5 y 8

1 1

..C:mpatible w/ Movable In-Cr.re v es Yes Deter. tor System Fuel Tube Materiai Zircsloy-4 Zircaloy-4 Fuel Red Clad CDs in. 0.422 0.424 l Fuel Rod Clad Thickness, in. 0.0243 0.030

[ Fuel / Clad Gap, mii 7.5 7.5 Fuel Pellet dia., in. 0.3659 0.3565 Guide Thimble Material Zircaloy-4 Zircaloy-4 G'uide Thimble ID, in.* 0.499 0.511 Structural Mat'l-Fiva laner*

Zircaloy-4 Zircaloy-4 Straps Grids Inconel Springs

• Above dashpot 23 L