ML20079S241

Provides 10CFR50.46 Annual Rept-ECCS Evaluation Model Revisions

ML20079S241
Person / Time
Site:	Callaway
Issue date:	10/19/1994
From:	Schnell D UNION ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
ULNRC-3087, NUDOCS 9410270078
Download: ML20079S241 (19)

Text

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1901 Chouteau Avenue Post Othee Bcx 149 St Lox. Mssaan EJIEE 314 554 2650 '

Donald F. Schnell 01%T sen,artueerevsen Etzernic October 19, 19c4 &c.

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U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137 Washington, DC 20555 Gentlemen: ULNRC-3087 t

DOCKET NUMBER 50-483 -

CALLAWAY PLANT 10CFR50.46 ANNUAL REPORT-ECCS EVALUATION MODEL REVISIONS Ref erences : 1) ULNRC-2141 dated 1-19-90

! 2) ULNRC-2373 dated 2-28-91 l

3) ULNRC-2439 dated 7-19-91 1

4) ULNRC-2664 dated 7-16-92 l

5) ULNRC-2822 dated 7-15-93

6) ULNRC-2892 dated 10-22-93

7) Westinghouse letter NTD-NRC-94-4306 dated 9-21-94 l Attachment 1 to this letter describes changes j to Westinghouse ECCS Evaluation Models which have been j implemented for Callaway for the time period from l

October 1993 to October 1994. Attachment 2 provides an j ECCS Evaluation Model Margin Assessment which accounts l for the peak cladding temperature (PCT) changes

resulting from the resolution of the issues described l

in Attachment 1 as they apply to Callaway. References f 1-6 above transmitted prior 10CFR50.46 reports.

l l Attachment 1 describes the resolution of those issues which have been implemented for Callaway.  !

j The margin allocations for Callaway to date are l identified in Attachment 2. Westinghouse will soon conclude their evaluation of the SBLOCTA Code Axial Nodalization issue discussed in Reference 7. We

, anticipate approximately 425 F of net, reduction in l SBLOCA PCT and will submit a 30-day report after being i notified by Westinghouse. Since the PCT values i determined in the large and small break LOCA

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9410270078 941019 'I PDR ADOCK 05000483 . Q. l p PDR i

U.S. Nuclear Regulatory Commission Page 2 l

analyses of record, when combined with all PCT margin allocations, remain well below the 2200 F regulatory limit, no reanalysis is planned by Union Electric.

Should you have any questions regarding this l letter, please contact us.

. Ver trul ours, 1

l l d l Donald F. Schnell GGY/jdg l

Attachments l

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I cc: 'T. A. Baxter, Esq.

Shaw, Pittman, Potts & Trowbridge 2300 N. Street, N.W.

Washington, D.C. 20037 i

M. H. Fletcher Professional Nuclear Consulting, Inc.

18225-A Flower Hill Way i Gaithersburg, MD 20879-5334 i L. Robert Greger  ;

} Chief, Reactor Project Branch 1 '

U.S. Nuclear Regulatory Commission Region III j 801 Warrenville Road Lisle, IL 60532-4351 Bruce Burtlett Callaway Resident Office l

, U.S. Regulatory Commission I l RR#1 1 Steedman, MO 65077 L. R. Wharton (2)

! Office of Nuclear Reactor Regulation l U.S. Nuclear Regulatory Commission l 1 White Flint, North, Mail Stop 13E21

, 11555 Rockville Pike

.>ckville, MD 20852 i

Manager, Electric Department

Missouri Public Service Commission j P.O. Box 360 j Jefferson City, MO 65102 I ,

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ATTACHMENT ONE l

CHANGES TO THE WESTINGHOUSE 5

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1 ECCS EVALUATION MODELS ,

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e 2-ECCS EVALUATION MODEL CHANGES AND CORRECTIONS

1. Vessel and Steam Generator Calculation Corrections in LUCIFER *

2. ISHII Drift Flux Correction

3. NOTRUMP Point Kinetics Correction

4. Core Node Initialization Correction

5. NOTRUMP Heat Link Pointer Correction l

6. Fuel Rod Model Corrections for SBLOCA i

7. Large Break LOCA Fuel Rod Model Corrections l

8. High Temperature Fuel Rod Burst Model '

9. Large Break LOCA Rod Internal Pressure Issues

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10. Hot Assembly Average Rod Burst Effects * ,

11. Revised Burst Strain Limit Model* l l

12. Small Break LOCA Limiting Time in Life *  ;

13. NOTRUMP Boiling Heat Transfer Correlation Corrections *

14. NOTRUMP Steam Line Isolation Logic Corrections *

15. NOTRUMP Core Node Zire Oxide Initialization Correction l

16. Charging / Safety Injection System Issues

17. Double-Disk Gate Valve Pressure Equalization

• Results in PCT sllocation in Attachment 2

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l. VESSEL AND STEAM GENERATOR CALCULATION CORRECTIONS IN LUCIFER The LUCIFER code is used to generate the component databases, from raw input 1 i data, to be used in the small and large break LOCA analyses. Errors were found in the VESCA.L subroutine of the LUCIFER code. These errors were in the geometric and mass calculations of the vessel and steam generator portions of the needed data. All LOCA analyses using the LUCIFER code outputs are affected by these error corrections. The errors were corrected in a manner to maintain the consistency of the LUCIFER code.

The errors were determined to be a Nondiscretionary Change as described in Section 4.1.2 of WCAP-13451 and were corrected in accordance with Section 4.1.3 of WCAP-13451. Representative plant calculatiora indicate a net PCT benefit of-16 F for small break LOCA and a benefit of-6 F for large break LOCA. ,

2. ISIIII DRIFT FLUX CORRECTION An error was discovered both in WCAP-10079-P-A, "NOTRUMP- A Nodal Transient Small Break and General Network Code," and the relevant coding in NOTRUMP SUBROUTINE ISHIIA which led to an incorrect calculation of the drift flux in NOTRUMP when a laminar film annular flow was predicted. The affected equation in WCAP-10079-P-A is Equation G-74 wherein a factor of 'g',

the gravitational constant, was inadvertently omitted from both the documentation and the equivalent coding. The correction of this error returned NOTRUMP to consistency with the ultimate reference for the affected correlation.

This was determined to be a Nondiscretionary Change as described in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of

. WCAP-13451. Representative plant analyses were used to estimate a generic PCT effect of 0"F for small break LOCA.

3. NOTRUMIUHINT KINETICS CORRECTION An error was discovered in the coding used in the NOTRUMP User External SUBROUTINE VOLHEAT. The coding did not correctly perform the calculation described by Equation 3-12-28 of WCAP-10054-P-A, " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code." This calculation is only used during the time when the point kinetics option is used to determina the core power before reactor trip. Therefore, any analysis which used the more conservative assumption of constant core power until reactor trip time is not affected by this error. The correction of this error returned NOTRUMP to consistency with WCAP-10054-P-A.

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This was detennined to be a Nondiscretionary Change as described in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of WCAP-13451. Representative plant analyses were used to estimate a generic PCT effect of 0 F for small break LOCA.

4. CORE NODE INITIALIZATION CORRECTION An error was discovered in how the properties of CORE NODE components were initialized for non-existent regions in the adjoining FLUID NODE. In particular this led to artificially high core temperatures during the time step when the core mixture level crossed a node boundary, conservatively causing slightly more core mixture level depression than appropriate during this time step. Correction of this error allows for a smoother mixture level uncovery transient during node crossings.

This was detennined to be a Nondiscretionary Change as described in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of WCAP-13451. The nature of this error led to an estimated generic PCT effect of 0 F for small break LOCA.

5. NOTRUMP IIEAT LINK POINTER CORRECTION l

An error was discovered in how NOTRUMP initialized ceitain HEAT LINK pointer variables at the start of a calculation. Correction of this error returned NOTRUMP to consistency with the original intent of this section of coding. ,

This was detennined to be a Nondiscretionary Change as described in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of WCAP-13451. Representative plant analyses were used to estimate a generic PCT effect of 0 F for small break LOCA.

6. FUEL ROD MODEL CORRECTIONS FOR SBLOCA A number of minor programming errors were corrected in the fuel rod heatup code used in SBLOCA analyses. These corrections were related to:

1. Individual rod plenum temperatures j 2. Individual rod stack lengths i 3. Clad thinning logic

4. Pellet / clad contact logic

5. Corrected gamma redistribution j 6. Including ZrO 2thickness at t=0 initialization

7. Numerics and convergence criteria of initialization

These changes were determined to be Nondiscretionary Changes as described in Section 4.1.2 of WCAP-13451 and were implemented in accordance with Section 4.1.3 of WCAP-13451. The cumulative effect of the error corrections and convergence criteria change was found to be less than approximately 4 F. This change is therefore judged to have a negligible effect on PCT and on a generic )

basis the estimated effect is reported as 0 F for small break LOCA.

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7. LARGE BREAK LOCA FUEL ROD MODEL CORRECTIONS l

Minor errors in the rod heatup code used in large break LOCA analyses were i corrected. These errors concemed conditions which exist during periods of

pellet / clad contact and the internal bookkeeping logic associated with clad thinning.

These changes were determined to be Nondiscretionary Changes as described in Section 4.1.2 of WCAP-13451 and were implemented in accordance with Section 4.1.3 of WCAP-13451.

Representative plant calculations have shown that these corrections have a negligible effect on PCT for near Beginning-of-Life (BOL) fuel rod conditions (i.e. < 2000 MWD /MTU). These effects become prevalent as burnup increases, but are not expected to be of any significance until pellet / clad contact is predicted for steady-state operating conditions (typically > 8000 MWD /MTU). These corrections therefore result in a negligible PCT impact for the large break LOCA licensing basis PCT which is calculated with near BOL conditions. This impact is reported generically as O'F for large break LOCA.

8. HIGII TEMPERATURE FUEL ROD BURST MODEL A model for calculating the prediction of zircaloy cladding burst behavior above the previous limit of 1742 F was implemented. This model was described to the NRC in:

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I.ctter ET-NRC-92-3746, N. J. Liparulo (W) to R.. C. Jones (NRC),

" Extension of NUREG-0630 Fuel Rod Burst Strain and Assembly Blockage Models to High Fuel Rod Burst Temperatures",

September 16,1992.

This was determined to be a Nondiscretionary Change as described in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of WCAP-13451. This issue does not result in the assessment of any penalties to the Callaway large break LOCA analysis since burst was predicted to occur below 1742 F and the new model does not apply. See the related Issue 9 below.

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9. LARf BREAK LOCA ROD INTERNAL PRESSURE ISSUES Westinghouse recently completed an evaluation of a potential issue conceming the impact of increased beginning of life rod internal pressure (RIP) uncenainties on

LOCA analyses. Historically, befnning of life fuel pressure and tempemture uncertainties were based upon enu m iife considerations. These RIP uncertainties

, were found to be potentially nonconservative. During the evaluation of this issue, j a second issue related to the applicability of generic Integral Fuel Burnable Absorber (IFBA) fuel analyses to updated LOCA Evaluation Models was also 4

identified and combined with this issue since the underlying mechanisms were the same.

4 The technical evaluation of this issue concluded that both the RIP uncertainty and the current IFBA designs with 200 psig initial fill pressure fuel typically will result in a maximum 15 F PCT variation. Consequently, RIP manufacturing uncertainties and 200 psig initial fill pressure IFBA fuel do not have significant effects on the large break LOCA analyses. Also, based on these results, it was concluded that only nominal RIP (with an upper bound bias) should be used in the LOCA analyses for fuel designs with an initial cold fill pressure 2 200 psig. This is consistent with past LOCA analyses.

Specific analyses were perfonned for all plants with initial fill pressure < 200 psig. A +49 F PCT penalty was assessed for the Callaway LBLOCA analysis for 100 psig IFBA fuel, offset by a -89 F PCT benefit from a Callaway-specific reanalysis of the limiting large break clad temperature transient using the revised LOCBART code, as discussed in ULNRC-2892 dated 10-22-93.

10. HOT ASSEPBLY AVERAGE ROD BURST EFFECTS The rod heatup code used in small break LOCA calculations contains a model to calculate the amount of clad strain that accompanies rod burst. However, the methodology which has historically been used did not apply this burst strain model to the hot assembly average rod. This was done so as to minimize the rod gap and therefore maximize the heat transferred to the fluid channel, which in turn would maximize the hot rod temperature. However, due to mechanisms j governing the zirc-water temperature excursion (which is the subject of the SBLOCA Limiting Time-in-Life penalty for the hot rod), modeling of clad burst strain for the hot a: sembly average rod can result in a penalty for the hot rod by increasing the channel enthalpy at the time of PCT. Therefore, the methodology has been revised such that burst strain will also be modeled on the hot assembly i avemge rod. i l

This was determined to be a Nondiscretionary Change as described in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of WCAP-13451.

Representative plant calculations have shown that this change introduces an approximate 10% increase in the SBLOCA Limiting Time-in-Life penalty on the hot rod (+2"F for Callaway). However, this penalty is offset by the Revised Burst Strain Limit Model described in Issue 11 below. See also the related Issue 12 below.

I1. REVISED BURST STRAIN LIMIT MODEL A revised burst strain limit model which limits strains is being implemented into the rod heatup codes used in both large break and small break LOCA. This raodel, which is identical to that previously approved for use for Appendix K analyses of Upper Plenum Injection plants with WCOBRA/ TRAC, as described in WCAP-10924-P-A, Rev.1, Vol.1, Add. 4, " Westinghouse Large Break LOCA Best Estimate Methodology: Volume 1: Model Description and Validation, l

. Addendum 4: Model Revisions," 1991.

This has been determined to be a Nondiscretionary Change as discussed in Section 4.1.2 of WCAP-13451 and is being implemented in accordance with Section 4.1.3 of WCAP-13451.

The estimated effect on large break LOCA PCT ranges from negligible to a moderate, unquantified benefit which will be inherent in calculations once this model is implemented. In small break LOCA, representative plant calculations indicate that the magnitude of the benefit (-2 F for Callaway) is conservatively estimated to be exactly offsetting to the penalty introduced by the Hot Assembly Average Rod Burst Effects discussed in Issue 10 above. See also the related Issue 12 below.

12. SMALL BREAK LOCA LIMITING TIME IN LIFE-ZIRC/ WATER OXIDATION TEMPERATURE EXCURSION Westinghouse recently completed an evaluation of a potential issue with regard to burst / blockage modeling in the Westinghouse small break LOCA evaluation model. This potential issae involved a number of synergistic effects, all related to the manner in which the small break model accounts for the swelling and burst of fuel rods, modeling of the rod burst strain, and resulting effects on clad temperature and oxidation from the metal / water reaction models and channri blockage.

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Fuel rod burst during the course of a small break LOCA analysis was fourJ to potentially result in a significant temperature excursion above the clad temperature transient for a non-burst case. Since the methodology for SBLOCA analyses had been to perform the analyses at a near beginning of life (BOL) condition, where rod internal pressures are relatively low, most analyses did not result in the occurrence of rod burst, and therefore may not have reDected the most limiting time in life PCT. In order to evaluate the effects of this phenomenon, Westinghouse has developed an analytical model which allows the prediction of rod burst PCT effects based upon the existing analysis of record.

The mutually exclusive penalties discussed in Attachment 2 of ULNRC-2822 dated 7-15-93, small break Burst and Blockage penalty and BOL Rod Intemal Pressure (RIP) Uncenainty penalty, have been resolved. At the time ULNRC-2822 was submitted, both of these penalties were 20 F and a commitment was made to track and repon whichever was higher. Since that time, Westinghouse has determined that the BOL RIP Uncertainty is no longer a safety concern with respect to the 10CFR50.46 Acceptance Criteria and will no longer be tracked since it applies only to PCTs below approximately 1700 F where burst is not a concern. As such, only a Burst and Blockage penalty, which is a function of the base PCT plus margin allocations, will be tracked. Given the 16 F benefit discussed in Issue 1 above, the previous 20 F penalty is reduced to 15 F. Since the Burst and Blockage margin allocation is not a fixed value, it will be tracked separately.

13. NOTRUMP BOILING HEAT TRANSFER CORRELATION CORRECTIONS This closely related set of errors deals with how the mixture velocity is defined for use in various boiling heat transfer regime correlations. The previous definition for mixture velocity did not properly account for drift and slip effects calculated in NOTRUMP. This error particularly affected NOTRUMP calculations of heat transfer coefHeient when using the Westinghouse Transition Boiling Correlation and the Dougall-Rohsenow Saturated Film Boiling Correlation.

In addition, a minor typographical error was also corrected in the Westinghouse Transition Boiling Correlation.

This was determined to be a Nondiscretionary Change as described in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of WCAP-13451. Representative plant calculations for this issue resulted in an l estimated PCT benefit of -6 F for small break LOCA.

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14. NOTRUMP STEAM LINE ISOLATION LOGIC CORRECTIONS This error consists of two ponions: a possible plant-specific effect which only applies to recent NOTRUMP analyses which model Main Feedwater Isolation (FWI) on a safety injection (SI) signal, and a generic effect applying to all previous analyses.

The possible plant-specific effect was the result of incorrect logic which caused main steam line isolation to occur on the same signal as FWI. Therefore, when the SI signal was chosen through user input to be the appropriate signal for FWI, it also caused the steam line isolation to occur on an SI signal. This is inc :.sistent with the standard conservative assumption of steam line isolation on Loss of Offsite Power coincident with the earlier Reactor Trip signal.

The generic effect was the result of incorrect logic which always led to ti e isolation functions occurring at a slightly later time than when the appropriate signal was generated. l l

This was determined to be a Nondiscretionary Change as described in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of i WCAP-13451.

l Representative plant calculations for this issue resulted in the assessment of an estimated PCT penalty of + 18 F for small break LOCA for the generic ponion of the issue. The plant-specific penalty of +12 F does not apply to Callaway because this issue applies only to NOTRUMP analyses more recent than Callaway's.

15. NOTRUMP CORE NODE ZIRC OXIDE INITIALIZATION CORRECTION 1

NOTRUMP models two regions for each core node analogous to the two (mixture  !

and vapor) regions in adjoining fluid nodes. During the course of a transient, NOTRUMP tracks region-specific quantities for each core node. Erroneous logic caused incorrect initialization of the region-specific, fuel cladding zirc oxide thickness at times prior to the acttial creation of the relevant region during the core boil-off transient.

This was determined to be a Nondiscretionary Change as descritxxl in Section 4.1.2 of WCAP-13451 and was corrected in accordance with Section 4.1.3 of WCAP-13451. Representative plant calculations led to an estimated generic PCT effect of 0 F for small break LOCA. I

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16. CIIARGING/ SAFETY INJECTION SYS l'EM ISSUES Westinghouse has recently completed its evaluation of a potential safety issue regarding four specific issues related to the design and use of the miniflow line l for the centrifugal charging pumps. Issue 1 involves the operation of the i centrifugal charging pump (CCP) miniflow line during accident conditions. A CCP runout condition may occur if the injection lines were balanced with the miniflow path closed and credit was taken for operator action to isolate the

! miniflow line during the accident. Also, the existence of this condition may impact the plant-specific small break LOCA analysis. Issue 2 is concerned with l the relationship between the plant-specific ECCS analysis and the plant conGguration. Many utilities have made changes to the configuration of the ECCS. As a result, the utility should have revised the ECCS flow calculations to i be consistent with the ECCS configumtion. If the ECCS Gow calculation was not l revised to be consistent with the ECCS configuration, then the plant may be

operating in an unanalyzed condition. Issue 3 involves miniflow oriGce plates that are used for the CCPs. Westinghouse has supplied two Jifferent orifice types to utilities for the charging pump miniflow line. One type a designed to provide 60 gpm at a differential head of 6000 feet. The other type is designed to provide a maximum of 70 gpm at a differential head of 6000 feet. Additional confinnation testing indicates that the orifice plates will allow a higher than design flow mte through the orifice at the design differential head. As a result, a discrepancy may exist between the installed miniflow line capacity and the ECCS analysis assumptions. The discrepancy would occur if the ECCS analysis assumed that the miniflow line resistance was based on the orifice allowing design flow at the design head as opposed to the higher as-tested flow and head. Issue 4 involves those utilities that made changes to the design bases for the charging pump miniflow isolation valves. While making these changes, the utility should have reviewed, among ether things, its specific licensing commitments for power lock-out, since closure of one isolation valve may affect more than one ECCS subsystem. If the utility made changes to the design bases and did not review its licensing commitments for power lock-out, then the utility may be operating outside its licensing bases.

The Callaway design has always provided an interlock for the normally open CCP miniflow valves (A train BG-HV-8110 and B train BG-HV-8111) to close on an SI signal coincident with high flow through the Boron Injection Tank (greater than 258.9 gpm). Low flow (less than 173.5 gpm) through the BIT opens or re-opens the valves coincident with an SI signal. This design is operator-independent and occurs through automatic signals. As such, the Grst issue is not applicable to Callaway since our miniflow valves are designed to automatically ,

change position based upon flow through the BIT.

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Since the interlock with BIT flow for the miniflow valves was part of the original plant design, the ECCS analysis of record incorporates the automatic opemtion of these valves. Therefore, the second issue is not applicable to Callaway. The high head ponion of the ECCS is balanced with these valves closed, consistent with the ECCS analysis.

The third issue involves the design of the miniflow ori6ces provided by Westinghouse for the CCPs. Additional testing has detennined that the orifice may pass up to 67 gpm for the orifice design at Callaway. Westinghouse has assumed a flow rate of 60 gpm ( 2%) for these ori6ces. Westinghouse has determined that the impact of the additional flow is acceptable for Callaway.

The fourth issue concems spurious operation of the miniflow valves. Inadvenent closure of the miniflow valve could result in failure of the CCP. FSAR Table 6.3-5 discusses failure of the miniflow valves to open or to close and the effects upon single failure should this event occur. The design basis of Callaway accounts for this concern.

As such, no PCT penalties are assessed against small break LOCA for Callaway.

17. DOUBLE-DISK GATE VALVE PRESSURE EOUALIZATION Westinghouse completed the evaluation of a potential issue concerning use of double-disk gate valves in the emergency core cooling system (ECCS) as hot leg isolation valves (e.g., EJ-HV-8840, EM-HV-8802A,B). Use of these double-disk gate valves may involve an inner disk pressure equalization line that could set up a leak path into the hot leg during cold leg injection following a loss of coolant accident (LOCA). This condition could lead to inadequate cold leg injection resulting in an increase in PCT.

The design characteristic of a double-disk gate valve provides isolation by the downstream disk sealing against the valve seat. The mechanical seat;ng force and the hydraulic force from the upstream (SI pump) act to provide force to the valve seal surfaces. The double-disk gate valve design results in a volume of fluid which is enclosed between the disks when the valve is closed. As the fluid volume heats up, pressure greater than system pressure may develop and may ,

cause the disks to bind against the seats to the extent that the valves cannot bc l

opened. To avoid this, many double-disk gate valves have been modified to include a pressure equalization line or a small hole in one of the disks to relieve the pressure between the disks. Based on generic leakage calculations it was determined that the double-disk gate valves modified to eliminate concerns for thermal binding could leak as much as 30 gpm per valve. This leakage into the RCS hot legs will increase steam binding during reflood ann result in an increase in the calculated peak cladding temperature.

This issue only affects plants that have modified the configuration of the double-disk gate valves that could be susceptible to pressure locking and/or thermal binding. A review of the current ECCS configuration at Callaway indicates that l the double-dis.k gate valves installed in the ECCS do not use a pressure equalization device, e.g., a bypass line or a relief hole drilled in one of the disks.

Therefore, this concem is not applicable to Callaway at this time and no large break LOCA PCT penalty is assessed. This issue is still under investigation generically by the industry.

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MARGIN ASSESSMENT FOR CALLAWAY 4

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i LARGE IIREAK LOCA A. ANALYSIS OF RECORD PCT = 2014 F B. 1989 LOCA MODEL ASSESSMENTS + 10 F (refer to ULNRC-2141 dated 1-19-90)

C. 1990 LOCA MODEL ASSESSMENTS +0F (refer to ULNRC-2373 dated 2-28-91) 1 D. 1991 LOCA MODEL ASSESSMENTS + 10 F (refer to ULNRC-2439 dated 7-19-91)

E. 1992 LOCA MODEL ASSESSMENTS, MARGIN + 29 F ALLOCATIONS, AND SAFETY EVALUATIONS (refer to ULNRC-2664 dated 7-16-92 and ULNRC-2892 dated 10-22-93)

F. 1993 LOCA MODEL ASSESSMENTS -

65"F (refer to ULNRC-2822 dated 7-15-93 and ULNRC-2892 dated 10-22-93)

G. CURRENT LOCA MODEL ASSESSMFETS -

OCTOBER 1994

1. LUCIFER ERROR CORRECTIONS -

6F (see Issue 1 of Attachment 1)

2. POWER SHAPE SENSITIVITY MODEL +0F (PSSM)

(refer to Item 5 of Attachment 1 to ULNRC-2822 dated 7-15-93)

LICENSING BASIS PCT + MARGIN ALLOCATIONS = 1992 F ABSOLUTE MAGNITUDE OF MARGIN ALLOCATIONS SINCE LAST 30-DAY REPORT (ULNRC-2892) = 6F i

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SMALL BREAK LOCA A. ANALYSIS OF RECORD PCT = 1528 F B. 1989 LOCA MODEL ASSESSMENTS + 229 F (refer to ULNRC-2141 dated 1-19-90)

C. 1990 LOCA MODEL ASSESSMENTS +0F I

(refer to ULNRC-2373 dated 2-28-91)

D. 1991 LOCA MODEL ASSESSMENTS + 57 FI (refer to ULNRC-2439 dated 7-19-91) l E. 1992 LOCA MODEL ASSESSMENTS AND +0F SAFETY EVALUATIONS (refer to ULNRC-2664 dated 7-16-92)

F. 1993 LOCA MODEL ASSESSMENTS -

13 F2 (refer to ULNRC-2892 dated 10-22-93)

G. 1993 SAFETY EVALUATIONS +4F3 (refer to ULNRC-2822 dated 7-15-93) l H. BURST AND BLOCKAGE / TIME IN LIFE + 15 Fl (This PCT assessment is tracked separately since it will change depending on future margin allocations.)

I. CURRENT LOCA MODEL ASSESSMENTS -

OCTOBER 1994

1. LUCIFER ERROR CORRECTIONS -

16 F (see Issue 1 of Attachment 1)

2. NOTRUMP BOILING HEAT TRANSFER -

6F CORRELATION ERROR (see Issue 13 of Attachment 1)

3. NOTRUMP STEAM LINE ISOLATION LOGIC + 18 F ER.ROR (see Issue 14 of Attachment 1)

LICENSING BASIS PCT + MARGIN ALLOCATIONS = 1816 F ABSOLUTE MAGNITUDE OF MARGIN ALLOCATIONS SINCE LAST 30-DAY REPORT (ULNRC-2892) = 40'F

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4 NOTES:

1. See Issue 12 of Attachment 1. The Small Break Burst and Blockage penalty is a function of the base PCT plus margin allocations and is being tracked separately in this report.

2. At the January 12,1994 meeting between Westinghouse and the NRC, Westinghouse agreed to provide to the NRC an addendum to WCAP-10054-P-A describing the SI model used in NOTRUMP including SI to the broken loop.

Addendum 2 to WCAP-10054 has been submitted to NRC. It references the improved condensation model (COSI) described in WCAP-11767 and provides justification for application of this model to small break LOCA calculations. In the interim, Union Electric will track the Peak Cladding Temperature (PCT) change reported in ULNRC-2892 (+ 150 F/-150'F) as a permanent change to Callaway's calculated PCT.

3. The +4.0 F Cycle 6 CRUD Deposition penalty will be carried until such time as it is evaluated to no longer apply.

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