ML17199G033

Safety Evaluation Supporting Amend 95 to License DPR-19

ML17199G033
Person / Time
Site:	Dresden
Issue date:	03/31/1987
From:	Office of Nuclear Reactor Regulation
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ML17199G035	List: ML17199G033 ML17252A644
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NUDOCS 8704060433
Download: ML17199G033 (11)

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 95 TO PROVISIONAL OPERATING LICENSE NO. DPR-19 COMMONWEALTH EDISON COMPANY DRESDEN NUCLEftR POWER STATION, UNIT NO. 2 DOCKET NO. 50-?37

1.0 INTRODUCTION

By letter dated December 10, 1986, fRef.1), as supplemented,Ja.nuary ?8, (Ref. ?fi) and February 5~ 1987 (Ref. 24), Commonwealth Edison Company (CECo) proposed to amend Provisional Operating License DPR-19 to support Cycle 11 operation of' Dresden Unit? with Exxon Nuclear Company (ENC) 9x9 reload fuel.

The proposed license and Technical specif'ication changes primarily involve:

(1) changes in nuclear limits to reflect the Cycle 11 9x9 reloa~* and supporting analyses; (2) incorporation of an expanded power/flow operating map; and (3) deletion of the license condition*for Single Loop Operation (SLO) and incorporation of SLO provisions in the body of the Technical Specifications.

The Februa\\~y 5, 1987 stipplement withdrew a portion of the December 10, 1986 amendment request and agreed to continue using the existing minimum Critical Power Ratio safety limit of 1.06 for residual 8x8R General Electric fuel in the Dresden 2 Cycle 11 reload in lieu of the originally proposed 1.05. This change did not affect the significant hazards consi~eration and therefore was not separately noticed.

In support of the D2Cll reload CECo submitted topical reports which described the reload analysis (Ref. 2) and the plant transient analysis (Ref. 3).

2.0 EVALUATION 2.1 Reload Description The Dresden 2 Cycle 11 (D2Cll) reload would incorporate a total of 172 unirradiated ENC XN-3 9x9 fuel bundles. This XN-3 batch is divided into 76 high Gadolinia bundles designated as XN-3H and 96 lbw Gadolinia bundles designated as XN-3L.

These bundles have a central region. enrichment of 3.35%

and 6 11 natural urania ends to yield an average as'sembly enrichment of 3.13%.

The remainder of the core is comprised of 412 previously irradiated

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• ENC-fabricated BxB assemblies, four 9x9 2-cyc1P irradiated demonstration assemhlies, and 13fi previously irradiated GE P8xP.R assemblies.

The D2C11 core loading would consist of the following fuel types:

Fuel Type Enrichment*

Number of Bundles Cvcle First Inserted GE P8x8R 2.65%

132 8

GE 8x8 LTA 2.65%

4 9

XN-1 8x8 2.83%

216 9

ENC 9x9 LTA 2.78%

4 9

XN-2 8x8

2. 83~b 196 10 XN-3H 3.13%

76 11 XN-3L 3.13%

96 11 724

• Bundle Average Enrichments The core would be operated under the Single Rod Sequencing fSRS) fuel management strategy to assure that the control rod withdrawal error will not be limiting.

2.2 Fuel Mechanical Design The mechanical design of the XN-3H and XN-3L 9x9 ~e1oad fuel is des~ribed in Reference 4.

The ENC and GE 8x8 fuel types to be returned to the Oresden-2 core were approved for operation in previous cycles~ *The prior ENC 8x8 fuel types carry the designations n~-1 and XN-2.

The 9x9 XN-3H and XN-3L fuel assemblies* a.re. identical with the except.ion of a

• chan~e in the number, locatior.,snd Gd2o3 enrichment* of* the Gadolinia-bearfog fuel rods *. XN-3H assemblies contain 8 Gadolinia rods at 4:.0 Wt % Gd2o3 Whereas XN-3L assemblies have 7 Gadolinia rods at 3.5 Wt % Gd203*

• Both fuel types contain 79 fuel rods (8 are tie rods) and ? water rods. Based on our previous review of the generic submittal (Ref. 4), we find the mechanical.,

design of the ENC 9x9 fuel for D2Cll reload is acceptable.

However, approval of extended exposure limits for future operating cyc1es is contingent on our approval of XN-NF-82-06~P), Revision 1, Supplement 1 (Ref. 5).

• 2.3 Nuclear Design The nuclear design for D2Cll has been performe~ with Exxon methodologies previously reviewed and approved (Ref. 6). The fuel loading pattern is give_n in Figure 4.3 of Reference 2.

The beginning of cycle shutdown margin (.SOM) is 4.34 percent delta Kand at minimum conditions is 1.2 percent delta-Ki well in excess of the required 0.38 percent delta K.

The standby ljquid control system (which is rlesigned to inject a quantity of boron solution that produces a concentration of no less than 600 ppm of boron in the.reactor core in less

1.'; *., /..... '.***, *. * : *. ** than 100 minutes) was calculated to provide a SDM of 5.4% delta K for cold conditions with all control rods in their full power positions. This fully meets shutdown requirements. Since these resuHs have been obtained by previously approved methods and fall within the expected range, we conclude that the nuclear design of the D2C11 reload core is acceptable.

For P.2C11 reload, there will be 10 ASEA-ATOM (A-A) control rods introduced into the single rod sequencing (SRS) locations of the core *. These control rncls hnve been generi ca 11 y reviewed and accepte~ by the NRC. (Ref. 7).

2.4 Th~rmal Hvdraulic Design The Exxon thermal-hydraulic methodolo~y and criteria used for the D2C11 design and analysis is the same as that used and approved in D2C10 reload.

The previous review concluded that hydraulic compatibility between GE and Exxon fuel is satisfactory and the calculation of core bypass flow and the safety limit minimum critical power ratio (SLMCPR) are acceptable.

The methodology for Cycle 11 is baset:' on ENC's revised critical power methodology in XN-NF-524, Revision 1 (Ref. 9) which incorporates a constant flow MCPR formulation for BWR applications. The staff has completed its generic review of XN-NF-524 and has concluded that the methodology for arriving at. f>'CPR safety limit is acceptable.

The XN-3 correlation used to develop the MCPR safety limit has been approved for application to both the resident rnr. and GE 8x8 fuel types (Ref. 10) and the new 9x9 fuel type (Ref~ 11). Based on our review of the D2Cll analysis (Ref. 2), we could not accept the licensee's proposed MCPR safety limit of 1.05 for all fuel types because this result does not represent the uncertainties of generic fuel specific parameters for GE fuel.

The licensee then proposed (Ref. ?.4) that the ~CPR safety limit for n2c11* remain at the value of 1.06 for GE P8x8R fuel based on previous ~RC approval (Ref. 25). Therefore, the MCPR value ofl.05 for all ENC fuel *"types and GE 8x8 fuel (Ref. ?Sl and that of 1.06 for GE P8x8R: in this reload are,

acceptable. The proposed operating limit MCPR for D2Cll i~ 1.32 for Exxon 9x9 f"uel and 1.31 for GE and Exxon 8x8 fuel. This includes. margin for application to future reloads of 0.02 CPR units for hoth ENC fuel and GE Bx.8 fuel and 0.01 for GE PBxBR fuel. This is a function of delta-WCPR for limiting transients which is discussed in Section 2.5 of this evaluation.

The therma1-hydraulic stability of the Cycle 11 core wa~ analyzed using the*

methods identified in XN-NF-80-19, Volume 4, Revision 1 (Ref. 12).

Ref~rence 12 cites the use of the COTRAN and COTRANSA 2 Models for use in the anf!.lysis of core thermal-hydraulic stability. The resultant maximum decay ratios for natural recirculation flow determined analytically using the approved COTRAN code at various power and flow conditions are 0.36 (47.63 rated power and 31.5% rated flow) at the 100% flow control line (FCL) and 0.58 (58% rated power) at the APRM rod block intercept.

For two pump minimum flow, the maximum decay ratio is 0.44 (67.4% rated power) at the APRM rod block line*.

Since all the decay ratios are less than the surveillance criterion of 0.75 a~ calculated by COTRAN, no stability technical specification surveillance is required for this Cycle 11 operation. However, the surveillance requirement for SLO is

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discussed in Section 2.7 of this evaluation. The Cycle 11 reload is the first full reload batch of ENC 9x9 fuel for Dresden 2.

Ori line stabilitv measurements at Susquehanna 2 and Grand Gulf 1 reactors have demonstrated that a single reload of ENC 9x9 fuel has little impact on the overall core stability. However, we are not at this time prepared to conclude that a full core loading of 9x9 fuel is acceptable from a stability standpoint. Therefore, prior to operation of Dres~en 2 with a full core of 9x9 fuel, the staf~ will consider the need to require on-line decay ratio measurement du~ing a future startup program.

2.5 Transient and Accident Analyses Core wide transients were analyzed with the COTRANSA computer code (Ref. 13 &

18) which includes a one-dimensional neutron kinetics model.for evaluation of the axial power shape response during transient events (the generator load rejection and feerlwater controller failure'. The referenced report has been reviewed by the staff and the methods for calculating the system transient response were found to be acceptable (Ref. 14 & 20).

Calculation of the chan9e in critical power ratio (CPR) during the core wide transient events involves the use of COTRANSA system results which serve as input to a hot channel analysis model (Ref. 15) used to calculate the delta-CPR values.-

Results of this hot channel model have been confirmed to be conservative (Ref. 15) by the licensee using the XCOBRA-T model calculation (Ref. 16) for D2Cll conditions. The XCOBRA-T model has been reviewed by the staff and found to be acceptable (Ref. 17). Hence the application of the COTRANSA results to Cycle 11 of Dresden Station Unit 2 is acceptable.

The licensee evaluaterl several categories of potential core wide transients for Cycle 11 and provided specific results for four transients, generator load rejection without bypass (LRw/oBP), turbine trip without bypass (TTw/oBP),

feedwater cnntro11er failure (FWCF) and less of feedwater heating (LQFWH).

The limiting transient is {dentified as the LRw/oBP.

The de1ta~WCPR for the LRw/oBP transient. was calculated using o. modified version of the COTRANSA.

delta-WCPR calculation model described in XN-NF-79-71(P) Rev. 2 (Ref. 18, 19 and 20), which has been re~iew~d and approved by the ~taff. At rated power; the COTRANSA delta-CPR was 0.24 (for ENC and GE 8x8 fuel) ancl 0.25 (for 9x9 'fuel) for LRw/flBP.

The most limiting event for reactor vessel nverpressu~1zation is the main steamline isolation valve (MSIV) closure without direct scram (sinale failure) on valve position. The maximum.value of the senserl pressure in the steam dame was 1309 psig which corresponds to a maximum vessel pressure of 1334 psig at the lower plenum.

These values are less than the.technical specification limit of 1345 psig as measured hy the steam dome pressure indicator and the 1375 psig ASME vessel pressure limit. This is acceptable.

.>_:...-.,.: The licensee has determined the required reduced f'low MCPR operating l.imit for off-ratecl*conditions to complement the Cycle 11 MCPR full flow operating limits during the automatic flow control (AFC) condition and in manual flow control (MFC).

The results are given in Tables 5.3, 5.4, 5.5, and 5.6 of Reference 3, and are acceptable.

The licenseP. also evaluated local transients, i.e., control rod withdrawal error (RWE), fuel loading error (FLE), and the ro~ drop acctdent (RDA), and a loss of coolant accident (LOCA), which are described as follows.

Using a rod block monitor (RBM) settir.g of 110 percent of full power for the.

RWE event results in a delta-WCPR of 0.13 for both 8x8 and 9x9 fuel. The limiting mislocated*assembly occurs when a fresh ENC 9x9 (XN~3) assembly is loaded into an identified core location and results in a delta-WCPR of 0.18 at a cycle exposure of 3,500 MWD/MTU.

These delta-WCPR values are bounded by the LRw/oBP transient event.

• The control rod drop accident evaluation yields a value of 125 cal/gm for the maximum deposited fuel rod enthalpy. This is well below the NRC required limit of 280 cal/gm, and is therefore acceptable..

The licensee h~s previously performed LOCA analyses (Ref. 21) which are ~al id for ENC 8x8 fuel (XN-1 and XN-2 reload) in Dresden 2. These analyses provided

.MAPLHGR limits for the ENC 8x8 fuel which remair. applicable in Cycle 11.

The licensee has also recently performed the limiting LOCA break calculation with a fol 1 core of ENC 9x9 fuel.

Two ca lcul at ions were* performed, one for full power and rated f1 ow ( 100/100) and the other at full pnwer and 87% fl ow (100/87), which is the minimum flow allowed fnr operation at -full power frC'm the extended load line limit analysis (ELLLA) region analyzed for D?Cll operation.

Both operating conditions resulted in essentially identical LOCA transients.

The power/flow of 100/87 resulted in slightly higher peak cladding temperature (PCT) and was used to calculate the LOCA-ECCS MAPLHGR limits. The final MAPLHGR calculation results are given in Table 1.1, Figure.*

1.1 and Table 3.2 of Reference 22.

The resulting PCT was ?.045° F at a burnup of 5,000 GWD/MTU, allowing a 155'° F margtn to the 10 CFR 50.46 limit (compared to 2,159° F for the 8x8 analysis)..

Metal water reaction also peaks at 2.44 percent at a burnup of 5,000 MWD/MTU remaining well below the 17 percert limit required by 10 CFR 50.46.

Since analysis of the LOCA was performed with reviewed and accepted methods (Ref.

8), and the results are well within the limits of 10 CFR 50.46, the staff finds the proposed MAPLHGR limit for D?.Cll acceptable.

2.6 Extended Load Line Limit Analysis The extended load line limit analysis (ELLLA) is an analysis to support plant normal operation in the region of power/flow map above the 100% power/100%

flow load line and bounded by the 108% APRM rod block line and the 100% rated power line. This added capability increases operating flexibility to permit flow compensation for xenon build-up following start-ups and for fuel depletion

• later in cycle, and to improve the efficiency of achieving and maintaining 100%

power.

Three off-rated power/flow conditions (100%/87%, 85%/61%, 67%/39%) on the expanded power/flow map were evaluated.

The results of the analysis (Ref.

21) indicate that all three points are bounded by the generator load rejection.

without bypass at the 100% power/100% flow condition on the operating map (Ref. 3).

Also, the change in slope of the APRM rod block line has been previously reviewed and approved.

The APRM rod block setting is Changed to provide a block at no greater than.58Wd + 50 power (where Wd is recirculation drive flow in percent).

This permits operation at 100 percent power down t6 87 percent flow.

For LOCA-ECCS concern, ENC has performed limiting LOCA break calculations in support of ENC-fuel in the Dresden Uriits fdr 100/87 and 100/100 conditions.

The results are described in Section 2.5 of this evaluation.

It is concluded that changes in core behavior caused by the extended operating range have been acceptably accounted fo~.

2.7 Single Loop Operation The licensee has proposed technical specifications to delete the Current license condition restricting single loop operation (SLO) to power levels less than 50 percent, and to incorporate all SLO provisions into the body* of the technical specifications.

The restrictions in the limitin~ conditions of operation (LCOs) include both dual loop and single loop operation technical specifications for average power range monitor (APRM) flux scram trip and rod block setting, an increase in the SLO safety limit MCPR value, and a *revision to the allowable ma~imum ~~erage planar linear heat generation rate (MAPLHGR) values for SLO.

The increase from.01 to.03 in the value of the MCPR safety limit and the reduction of the MAPLHGR multipliers from those values repo~ted in Reference 23 to 0. 70 for all GE fuel types for SLO have been previously approved for D3Cl0 reload.

This increase in SLMCPR for SLO is to account for core flow and tip

• reading uncertainties which are used in the statistical analysis of the safety limit.

The reduction factor (the MAPHGR multiplier) is to account for the more rapid loss of core flow during SLO.

The modification of the APRM scram trip and rod block setting is to account for reverse or stalled flow in the idle loop (the reduction is 3.5% for APRM scram trip setpoint and 4% for RBM setpoint for SLO).

The reduced flow operating limit curves for dual loop operation are applicable*

without modification for SLO because the reduced flow MCPR limit curves are based on a 2-pump event which bounds the one pump run-up.

However, the automatic flow control reduced flow MCPR curves do not apply in SLO because this mode of operation is prohibited at Dresden in SLO by the technical specification.

• Although core stability calculations for Dresden 2 Cycle 11 indicate suhstantial stability margins, the provisions of monitoring LPRM and APRM noise levels during SLO are included in the proposed technical specification changes to assure that should limit cycle oscillations occur, t.hey will be detected in a timely manner.

In addition, station procedures require monitoring of core plate delta-WP to assure jet pump operability.

Based on our review we have found that the licensee's proposal to remove the license condition for 50% power restriction (no longer needed to resolve sta.bility concerns) while in SLO and to incorporate the SLO provisions into.the body of the technical specifications is acceptable.

2.8 Technical Specification Changes To accomplish the proposed changes for SLO limits, incorporation of an expanded power/flow operating map, nuclear limits, the Bases and reference update, anc pages addition or removal, TS changes have been requested. These revisions are evaluated in Section 2.0 of this evaluation and are discussed as follows:

{1)

Specifications 1.1.A and 3.5.K.3, and Bases 1.1 and 3.6: The SLMCPQ of 1.05 for ENC fuel and GE 8x8 fuel, and that of 1.06 for GE P8x8R fuel are acceptable as previously approved values.

The increase of SLMCPR by 0.03 fnr SLO to account for the increased uncertainties in the total core flow and transversing incore probe reading is acceptable.

(2)

Specification 2.1.A.1:

APRM flux scram trip setting (run mode) is.58 Wd

+ 62 during dual loop operation and.58 Wd + 58.5 during single loop operation.

The 3.5% difference is due to the difference between dual loop operation and SLO drive +fow at thr:>. same core flow.

This is acceptable.

(:n Specification 2.1.B:

P.PRM Rod Rlock setting is 0.58 \\~d + 50 during dual loop operation and.58 Wd +46.5 during SLO.

The 3.53 difference is due to thP difference between dual loop operation and SLO drive flow at the same core flow.

This is acceptable.

(4)

Figure 2.1-3:

/l.n expanded power/flow operating map includipg APRM flow bias scram relationship to normal operating conditions has been added to the current TS.

This curve is based on ENC calculations and is acceptable.

(5) Table 3.2.3 and Basis 3.2: The APR~ trip setting and the rod block monitor setting are given for both dual loop operation and.SLO with 3.5%

and 4% difference, respectively, to account for reverse or stalled flow in the idle loop.

These revisions are acceptable *

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(6)

Specification 3.5.I and Pasis 3.6: Average Planar LHGR for SLO in Figure 3~5-1 shall be decreased to 70% of the original value to account for the assumption of an earlier occurrin9 boiling transiti-0n. This is acceptable.

(7)

Specification 3.5.J: This specification has been changed to add a reference to Figure 3.5-lA giving the linear hea.t generation rate (LHGR) curve for Exxon fuel.

The GE fuel LHGR remains at 13.4 KW/ft.

This is*

acceptable.

(8)

Specification 3.5K:

OLMCPR is 1.32 for Exxon 9x9 fuel 'and 1.31 for GE and Exxon 8x8 fuel. This is acceptahle.

(9)

Specification 3.5.K.2:

Figure 3.5-2 is added to account for the automatic flow control MCPR limit for reduced total cqre flow.

This is acceptable.

(10) Specification 3.5.K.3: The mean control rod scram inserti.on timP to 90%

ir:sertion ha.s been revised from ?.74 seconds for Cycle 10 to 2.77 seconds for Cycle 11 based on licensee's long-term measurement results.

• This is acceptable.

(111 Specification 3.6.3.and Basis 3.6: Procedures on SLO restrictions are specified. These are evaluated in Section 2.7. and are acceptable.

Figure 3. 6. 2 is added for therma 1 hydrau1 ic stability survei 11 ance in SLO

( therma 1 power versus *core f'l ow 1 imi ts).

(12) P.asis 3.6.H:

Recirculci.tion pump flow limitations are described in terms of a pump vibration due to a flow mismatch between the two sets of,iet pumps; reduced flow OLMCPR for automatic flow control (not applicable to SLO), and. stability surveillance for SLO.

These are ~ppropriatP..

3.0

SUMMARY

As a result of our review, which is described in Section 2.0 of this evaluation, we conclude that the proposed reload and technical specification changes are acceptable. This conclusion is based on the following:

1.

previously approved analysis methods and techniaues are employed,

2.

• the results of the transients and accidents which are affected by the reload are in conform~nce with our standard review criteria and the applicable regulations and, therefore, are acceptable for Cycle 11.

4. 0 ENVIRONMENTAL CONSIDERATION This amendment involves changes to requirements with respect to the instal-lation or use of facility components located within t~c restricted area as defined in 10 CFR Part 20 and changes to the surveillance requirements.

The staff has determined that the amendment involves no significant increase

1*

.. **.... *:.. 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 individual or cumulative occupational radiation exposure.

The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22<b) no environmental "impact*statement nor environmental assessment reed be prepared in connection with the issuance of this amendment.

5. 0 CONCLUSION The staff has concluded, based on the con~iderat1ons discussed above, that:.

(1) there is reasonable assurance that the health and safety of the public

• will rot be endangered by operation in the proposed manner, Rnd (2).such activities will be CC\\nducted in compliance with the Commissior.

1s regu1ations and the issuance of this amendment wi 11 not be i nimi ca 1 to the commC'n defense and security nor to the health ancl safety of the public.

Principal Contributor:

T. Huang l

Dated:

6.0 REFERENCES

1.

2.

3.

4.

5.

6.

7.

8.

Letter, C. M. Allen (CECo) to H. R. Oenton (NRC), dated December 10, 1986.

XN-NF-86-103, Dresden Unit 2 Cycle 11 Reload Analysis,. September 1986.

XN-NF..;86-102, Oresrlen Unit 2 Cycle 11 Plant Transient Analysis, September 1986.

X~-NF-86-67(P)(A), Revision 1, "Generic Mechar.ica~ Desigri for Exxon Nuc1 ear Jet Pump BWR Reload Fuel," September 1986.

XN-NF~82-06fP), Revision 1, Supplement 1, "Qualification of Exxon Nuclear Fuel for Extended Burnup, Supplement 1, Extended Burnup Qualification of ENC 9x9 RWR Fuel," April 1984.

• xN-NF-.80-19(P) (J!.), Volume 1 and XN-NF-80-19(P)(A), Volume 1, Supplements 1 & 2 "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic.

Methods for Design and Analysts," March 15, 1983.

TR-VR-85-225-A, "Topical Report of ASEA-ATOM BWR Control Blades for US

BWRs, 11 October 1985.

XN-NF-80-19(P) (A), Volume 2, "EN Methodology for Boi 1 ing Water Reactors EXEM BWR ECCS Evaluation Model~" September 1982.

).

9.

10.

lL 1 I')

13.

' '".. _. : ' -~ *.:~

~* ::

• .. \\ ____

,,..... XN-NF-524(P)(A), Revision 1, 11Exxon Nuclear Critical Power Methodology for BWRs, 11 November 1983.

Letter, H. Bernard (NRC) to G. F. Owsley (ENC),

11Acceptance for Referencing of Topical Report XN-NF-512, Revision l, 11 dated July 22, 1982.

Letter, C.O. Thomas (NRC) to J. C. Chandler (ENC),

11Acceptance for Referencin~ of Licensing Topical Report XN-NF-734, Confirmation of the XM-3 Critical Power Correlation for 9x9 Fuel Assernblies, 11 dated February 1, 1985.

XN-NF-80-19(P)(A), Volume 4, Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: J!pplication of the ENC Methodology to RWR Reloads," Fxxon Nuclear Company, September 1985.

XN-NF-80-19(P', Volume 3, Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors - THERMEX:

Thermal Limits Methodology, Summary Description," November 10, 1981.

14.

XN-~!F-80-19(P)(a), Volume 3, Revision 2, "Exxon Nuclear Methodology for Roiling Water Reactors - THERMEX:

Thermal Limits Methodology Summary Description 11, January 1987.

15.

XN-NF-85-113(P), "Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors," August 1986.

16~

XN-NF~84-105(P),

11 XCCRRA-T:

A Computer Code for BWR Transient Thermal Hydraulic Core Analysis, 11 May 31, 1985.

17. Letter G. C. Lainas to G. N. Ward (ENC),

11 Acc~ptance for Referencing of Licen~ing Topical Report XN-NF-84-105(P), "XCOBRA-T: *A Computer Code for BWR Transient Thermal Hydraulic Core Analysis, 11 October ?..7, 1986 *

. 18.

XN-NF-79-71(P), Revision 2, 11 ENC P1ant Transient Methodology for Boiling Water Reactors, 11 November 1981.

19.

XN-NF-79-7l(P), Revision 2, Supplement 3, "Revised Methodology for Including Uncertainties in Determining Operational Limits for Rapid Pressurization Transients in BWRs, 11 March 1985.

20.

XN-NF-79-71(P)(A) Revision 2, Supplemerts 1, 2 & 3, 11 Exxon Nuclear Transient Methodology for Boiling Water React0rs, 11 March 1986.

2L XN-NF-81-75(P),

11Dresden Unit 3 LOCA Analysis Using the ENC EXEM Evaluation Model 11, November 1981.

22.

XN-NF..;85-63(P), "Dresden Unit 3 LOCA-ECCS Analysis - MAPLHGR Results for 9x9 Fuel, 11 September 1985.

--~- '

7""' - 23.

NED0-24807(80NED031) Class 1~ "DrPsden Nuclear Power Station Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2 Single-Loop Operation," flecember 1980.

24.

Letter, I. M. Johnson (CECo) to H. R. Denton (NRC), dated February 5, 1987.

25.

Letter, J. D. Hegner (NRC) to L. DelGeorge (CECo), SER for D3C8 Reload, April 29, 1982.

26.

Letter I. M. Johnson (CECo) to J-!. R. Denton (NRC), datec! January 289 1987.

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