ML20045H974
| ML20045H974 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 07/15/1993 |
| From: | Schnell D UNION ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| ULNRC-2822, NUDOCS 9307220202 | |
| Download: ML20045H974 (13) | |
Text
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3 4 554 1660 Union July 15, 1993 f"
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. Etscruic wa 323 U.S. Nuclear Regulatory Commission Attn:
Document Control Desk Mail Station P1-137-Washington, DC 20555 Gentlemen:
ULNRC-2822 DOCKET NUMBER 50-483 CALLAWAY PLANT 10CFR50.46 THIRTY DAY REPORT-ECCS EVALUATION MODEL REVISIONS
References:
- 1) ULNRC-2141 dated 1-19-90
- 2) ULNRC-2373 dated.2-28-91
- 3) ULNRC-2439 dated 7-19-91
- 4) ULNRC-2664 dated 7-16-92
' to this letter describes changes to Westinghouse ECCS Evaluation Models which have been implemented for Callaway for the time period from June 1992 to June 1993. provides an ECCS Evaluation Model Margin Assessment'which accounts for the peak cladding temperature (PCT) changes resulting from the resolution of the issues described in as they apply;to Callaway.. References 1-4 above transmitted prior 10CFR50.46 reports.-
r describes the resolution of those
~
issues which have been implemented for Callaway.
The margin allocations for Callaway to date-are identified in Attachment 2.
Based on the criteria and reporting.
requirements of 10CFR50.46 (a) (3) (ii), as clarified in l
Section 5.1 of WCAP-13451 " Westinghouse Methodology for Implementation of 10CFR50.46 Reporting," the cumulative changes since the last 30-day report, Reference 3, are significant for large break LOCA and require another 30 day report.
However, since the PCT values. determined in the large and small break LOCA-analyses of record, when combined with all PCT margin allocations, remain well below the 2200*F regulatory limit, no reanalysis will be performed.
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9307220202 93071S PDR b
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UiS. Nuclear Regulatory Commission
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Should you have'any questions regarding this letter, please contact us.
Very truly yours, 4f un Donald F. Schnell 1
GGY/kea Attachments 1
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cc:
T. A.
Baxter, Esq.
Shaw, Pittman, Potts & Trowbridge 2300 N.
Street, N.W.
Washington, D.C.
20037 Dr. J.
O.
Cermak CFA, Inc.
18225-A Flower Hill Way Gaithersburg, MD 20879-5334-L. Robert Greger Chief, Reactor Project Branch 1 U.S.
Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 Bruce Bartlett Callaway Resident Office U.S.
Nuclear Regulatory Commission RR#1 Steedman, Missouri 65077 L.
R. Wharton (2)
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 1 White Flint, North, Mail Stop 13E21 11555 Rockville Pike Rockville, MD 20852 Manager, Electric Department Missouri Public Service Commission P.O.
Box 360 Jefferson City, MO 65102 l
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ULNRC-2 822 je ATTACHMENT ONE e
CHANGES AFFECTING CALLAWAY LARGE AND'
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SMALL BREAK LOCA PCT VALUES' h
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9 1.
AUXILIARY FEEDWATER FLOW TABLE ERROR The steam generator Auxiliary Feedwater (AFW) flowrate in the SBLOCA Evaluation Model (EM) is governed by the timing variable TIMESG(I).
A minor logic error associated with this variable was discovered which led to a step change in l
the AFW flowrate once the transient time passed the value of TIMESG(7).
Typically, this value is set equal to 11000 seconds and so this error would only affect very long transient calculations (much longer than any reported for Callaway in FSAR Section 15.6.5).
In addition, the nature of the error is to allow the AFW flowrate to immediatelv revert to the full value of the' Main Feedwater flowrate'.
This enormous step change has led to code aborts in the cases where it has occurred.
This logic was corrected as a Discretionary Change as described in Section 4.1.1 of WCAP-13451.
This determination was based on the fact that SBLOCA transients are generally terminated before the logic error can have an i
effect coupled with the code's inability to handle the step change if it does occur.
This error correction has no effect on any current or prior applications of the SBLOCA EM.
2.
STEAM GENERATOR SECONDARY SIDE MODELLING ENHANCEMENTS l
J set of related changes which make steam generator secondary side modelling more convenient for the user were implemented into NOTRUMP.
This model improvement involved several facets of feedwater flow modelling.
First, the-common donor boundary node,for the standard SBLOCA EM nodalization has been separated into two identical boundary i
nodes.
These donor nodes are used to set the feedwater enthalpy.
The common donor node configuration did not' allow i
for loop-specific enthalpy changeover times in cases where asymmetric AFW flowrates or purge volumes were being modeled for plant-specific sensitivities.
i The second improvement is the additional capability to initiate main feedwater isolation on either loss of'offsite r
power coincident with reactor trip (low pressurizer pressure) or alternatively on safety injection signal (low-low pressurizer pressure).
The previous model allowed this-function only on loss of offsite power coincident with reactor trip.
The auxiliary feedwater pumps are still assumed to start after a loss of offsite power with an appropriate delay time to model diesel generator start-up and bus loading times.
The final improvement is in the area of modelling the purging of high enthalpy main feedwater after auxiliary feedwater is calculated to start.
This was previously 1
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modelled through an approximate time delay necessary to purge the lines of the high.enthalpy main feedwater before credit could be taken for the much lower enthalpy-auxiliary feedwater reaching the steam generator. secondary.
This time delay _was a function of the plant-specific purge volume and j
the auxiliary feedwater flowrate.
The_new-modelling. allows the user to input the purge volume directly.
This then is used together with the. code calculated integrated feedwater-flow to determine the appropriate time at which the feedwater enthalpy can be assumed to change.
These improvements are considered to be a Discretionary Change as described in Section 4.1.1 of WCAP-13451.
Since they involve only enhancements to the capabilities and useability of the SBLOCA EM, and not changes to results calculated consistently with the previous model, these changes were implemented without prior review as discussed in Section 4.1.1 of WCAP-13451.
Because these enhancements only allow greater _ ease in modelling plant-specific steam generator secondary side behavior over the previous model, it is estimated that no effect will be seen in SBLOCA EM calculations.
3.
STRUCTURAL METAL HEAT MODELING A discrepancy was discovered during review of the finite element heat conduction model used in the WREFLOOD-INTERIM code of the LBLOCA EM to calculate heat transfer from scructural metal in the vessel during the reflood phase.
It-was noted that the material properties.available in the code corresponded to those of stainless steel.
While this is correct for the internal structures, it is inappropriate for the vessel wall which consists of carbon steel with a thin stainless internal clad.
This was defined as a Non-Discretionary Change per Section 4.1.2 of WCAP-13451, since there was thought to be potential for increased PCT with a more sophisticated composite model.
The'model was revised by replacing it with a more flexible one that allows detailed specification of structures.
The effect of this correction turned out to be a 25'F-PCT benefit.
4.
SPACER GRID HEAT TRANSFER ERROR IN BART During investigations into anomalous wetting and dryout behavior demonstrated by the BART grid model of the LBLOCA EM, a' programming logic error was discovered in the grid-heat transfer model.
The error caused the solution to be performed twice for each time step.
The error was traced back to the original coding used in all of the BART and LOCBART codes.
This was defined as a Non-Discretionary Change per Section 4.1.2 of WCAP-13451.
The error was
-2
i corrected, and a complete reverification'of the grid model.
was conducted and transmitted to the NRC (WCAP-10484, Addendum 1).
Calculations performed with the affected code have consistently demonstrated significantly.better grid wetting and lower clad temperatures.
A conservative estimate of zero degrees PCT penalty has been assigned for this issue.
5.
POWER SHAPE SENSITIVITY MODEL (PSSM)
Historically, chopped cosine power shapes have been assumed to produce limiting results in Westinghouse large break LOCA analyses.
However,'with the advent of more advanced models
(:BART and BASH) in the LBLOCA EM, it was discovered that under certain circumstances, top-skewed power shapes could potentially be more limiting.
The PSSM was developed to allow the' assessment of shape-specific PCT trends in large break LOCA.
As described in WCAP-12909-P and further clarified in ET-NRC-91-3633 (currently under NRC review),
the methodology was developed from a large database of large break LOCA analysis results which used a wide variety of full-power power shapes in typical twelve foot cores for Westinghouse supplied fuel.
This methodology change is considered to be a Non-Discretionary Change as described in Section 4.1.2 of WCAP-13451 and has been implemented prior to final NRC review in accordance with Section 4.1.3 of WCAP-13451.
The implementation of this methodole y reasonably assures j
that cycle-specific power distributions will not lead to results more limiting than those of the analysis of record.
Therefore, there is no PCT effect for this methodology.
6.
CRUD DEPOSITION Callaway has experienced deviations in axial offset
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A from core design predictions in Cycles 4, 5,
and 6.
The i
cores have behaved as predicted for the first six months of each cycle, but then begin to have a more negative axial offset than predicted.
l This axial offset indicates that the axial power distribution is becoming skewed toward the bottom of the core.
Westinghouse has reviewed operational data from these cycles and has concluded that these axial offset deviations appear to result from negative reactivity insertion in the top portion of the core during operation at Hot Full Power (HFP).
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Westinghouse has identified the accumulation of porous, boron-enriched CRUD deposits on the fuel in the upper half of the core as the probable cause of this axial offset anomaly.
Westinghouse has performed an assessment of the potential impact on LOCA analyses due to this CRUD deposition.
Westinghouse has reviewed the visual records of the Callaway fuel and has estimated the CRUD thickness as less than or equal to one mil.
The location and percent of surface area covered is defined as a non-uniform deposition primarily in the upper half of the core.
The CRUD deposition then smoothly transitions from the non-CRUD region to a region with CRUD.
The core CRUD is a porous agglomeration of particles that originates from the corrosion products released from the surfaces of the primary system.
The axial offset' anomaly is attributed to reaction mechanisns leading to enhanced boron concentrations within'these otherwise typical CRUD deposits.
Large Break LOCA The fuel rod conditions at the time of PCT were taken from the current analysis of record.
This-analysis was performed with a fuel rod peak burnup of less than 1000 MWD /MTU, at the burnup producing the maximum fuel rod average temperature or maximum stored energy.
CRUD deposition eventually results in a 1 mil thickness by the time a burnup of 4000 MWD /MTU is reached, producing a 10 F increase in the peak normal operation fuel average temperature.
In the burnup band of 1000 MWD /MTU to 4000 MWD /MTU, clad creep-down reduces the fuel rod average temperature of both IFBA and non-IFBA fuel by more than 145 F for norral. operation.
The 10 F increase in normal operation peak (hot spot) temperature due to CRUD is thus covered by clad creep-down effects with 135 F remaining to cover LOCA transient l
effects.
Based on the LOCA hot rod heat flux at the time:
and location of PCT (18,735 Btu /sq-ft hr) and the CRUD conductivity of 0.5 Btu /sq-ft/ F hr, the effect of 1 mil of CRUD on large break PCT is less than 6'F.
This_is more than offset by the 135'F reduction in fuel average temperature-at the start of the transient.
It is clear uhat the creep-down effects of the clad more than cover the effects of 1 mil of CRUD for the large break transient and no PCT penalty is required.
1 1
i Small Break LOCA
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The fuel rod conditions at the' time of' PCT were taken from j
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the current analysis of record.
In small break LOCA,. fuel i
rod heatup starts from the RCS system saturation temperature at the time of core uncovering, and the normal operation CRUD effect of 10*F on fuel rod temperature is.not applicable.
Based on the LOCA hot rod heat flux at the time.
and location of PCT (12,590 Btu /sq-f t hr) 'and' the CRUD
+
conductivity of 0.5 Btu /sq-ft/*F hr, the effect of 1 mil of CRUD on small break PCT is less than 4*F.
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ULNRC-2822 3
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ATTACHMENT TWO
.ECCS EVALUATION MODEL MARGIN ASSESSMEITI' FOR CALLAWAY l
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'LARGE BREAK LOCA A.
ANALYSIS OF RECORD PCT =-2014'F B.
1989 LOCA MODEL ASSESSMENTS
+
-10*F f
1 (refer to ULNRC-2141 dated 1-19-90)
C.
1990 LOCA MODEL ASSESSMENTS
+
0
- F.
(refer to ULNRC-2373 dated 2-28-91) i D.
1991 LOCA MODEL ASSESSMENTS
+-
10*F-(refer to ULNRC-2439 dated 7-19-91) 1 E.
1992 LOCA MODEL ASSESSMENTS, MARGIN
+ 31.7*F ALLOCATIONS, AND SAFETY EVALUATIONS
+
(refer to ULNRC-2664 dated 7-16-92)
F.
CURRENT LOCA MODEL ASSESSMENTS - JUNE 1993 25'F 1.
STRUCTURAL METAL HEAT MODELING - WREFLOOD CODE ERRORS (refer to Item 3 of Attachment 1) 6 2.
SPACER GRID HEAT TRAMSFER ERROR IN BART
+
0*F' (refer to Item 4 of Attachment 1) 3.
POWER SHAPE SENSlTIVITY MODEL (PSSM)
+
0*F.
i' (refer to Item 5 of Attachment 1)
G.
10CFR50.59 SAFETY EVALUATIONS - JUNE 1993
.l 1.
CRUD DEPOSITION
.+
0*F (refer to Item 6 of Attachment'1)
F LICENSING BASIS PCT + MARGIN ALLOCATIONS 2040.7*F
=
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NOTES:
s
- 1. The 1992 assessments included a LOCA Evaluation Model
- l penalty of +2*F for BOL Rod Internal Pressure Assumption, a LOCA-related margin allocation of +18.6*F for SG Flow Area Seismic /LOCA Tube Collapse, and-10CFR50.59 safety evaluation penalties of +10*F for Containment Purge
+
Effects and +1.1*F for reconstitution of fuel assembly-G87 (applicable only as long as G87 is in the core).
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8 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
+
0*F (refer to ULNRC-2373 dated 2-28-91)
]
1 D.
1991 LOCA MODEL ASSESSMENTS
+ 77'F (refer to ULNRC-2439 dated 7-19-91)
]
2 E.
1992 LOCA MODEL ASSESSMENTS AND SAFETY'
+0.1*F EVALUATIONS i
(refer to ULNRC-2664 dated 7-16-92) l F.
CURRENT LOCA MODEL ASSESSMENTS - JUNE 1993
- 1. AUXILIARY FEEDWATER FLOW TABLE ERROR
+
0*F (refer to Item 1 of Attachment 1) l 2.
STEAM GENERATOR SECONDARY SIDE MODELLING
+
0*F ENHANCEMENTS (refer to Item 2 of Attachment 1) i G.
10CFR50. 59 SAFETY EVALUATIONS - JUNE 1993 i
1.
CRUD DEPOSITION
+
4*F (refer to Item 6 of Attachment 1)
LICENSING BASIS PCT + MARGIN ALLOCATION = 1838.1*F l
NOTES
- 1.
The 1991 assessments originally included penalties of
-+37'F for Fuel Rod Model Revisions,'0*F for NOTRUMP Codef
.i Solution Convergence,.+40*F for SBLOCA Rod Internal i
Pressure Assumption, and 0*F for SBLOCA Broken Loop SI.
Flow Assumption. The SBLOCA Rod Internal Pressure penalty I
of +40*F was composed of two individual aspects. For SBLOCA analyses the limiting rod internal pressure (RIP) assumption depends on whether burst is predicted to occur. A higher RIP may lead to a higher calculated PCT if burst is predicted to occur. Conversely,.a lower RIP may decrease cladding creep (rod swell) away from the fuel pellets when the fuel rod internal pressure-is greater than the RCS pressure. Therefore, a lower RIP could then result in a higher calculated PCT, since.the cladding would be closer to the fuel pellet, for an analysis that did not predict fuel rod burst. Rod burst is not predicted to occur in the Callaway SBLOCA analysis of record (see FSAR Table 15.6-15). Therefore, to bound
L SMALL BREAK LOCA NOTES: (cont.)
1.
the effects of assuming a low initial RIP, a +20*F PCT penalty was assessed in 1991. The RIP issue also involved a Cladding Strain Model Error in the small break clad heatup calculation for which another +20*F PCT penalty was assessed in 1991, for a total ~of +40*F as reported in ULNRC-2439. Since the time of the 1991 report, a related issue, BOL Rod Internal Pressure Uncertainty, was opened for non-IFBA fuel rods. Using a conservative combination of BOL uncertainties resulted in an estimated decrease'of up to 65 psi in the predicted BOL RIP. Based on sensitivity analyses, a PCT penalty of +20*F was assessed in the 1992 50.46 report. Resolution of these issues as reported in ULNRC-2664 incorporated the Rod Internal Pressure Assumption portion of the original issue but not the Cladding Strain Model Error. As such, the original
+40*F PCT penalty reported in 1991 was reduced to +20*F for the RIP Assumption and a separate +20*F penalty was allocated for the RIP Uncertainty issue for a total of
+40*F as reported in 1992. The latter +20*F penalty for RIP Uncertainty applies only in the absence of rod burst.
RIP increases with burnup and a rupture of the cladding may be calculated for middle or end of life conditions. A rupture of the fuel rod could result in an increase in the PCT due to flow blockage effects. Therefore, for conservatism, PCT penalties for both cases, i.e. rod burst and absence of rod burst, will be addressed. If rod burst is postulated, a +20*F penalty for RIP Assumption applies as well as a penalty for Small Break Burst and Blockage. The penalty associated with Small Break Burst and Blockage is a function of the base PCT plus margin allocations and will increase / decrease as evaluations are performed. The determination of whether the RIP Uncertainty penalty of +20*F, applicable only in the absence of rod burst, or the Small Break Burst and Blockage penalty, +20*F for the current total PCT of 1838.l*F, is tracked and reported depends on which of these mutually exclusive penalties is higher. For the purposes of this report, neither penalty is greater'and no change to the overall +77'F penalty reported in 1991 is required.
2.
This penalty applies only as long as reconstituted fuel assembly G87 is in the core.
..... _ - _ _ _ -