ML20080H095

Forwards Addl Info Re Util Proficiency in Use of CESEC-III Computer Code,Per NRC 840126 Request.Encl Info Addresses Generic Ltr 83-11 & Demonstrates Validity of Cycle 8 to Cycle 6 Comparisons

ML20080H095
Person / Time
Site:	Fort Calhoun
Issue date:	02/08/1984
From:	William Jones OMAHA PUBLIC POWER DISTRICT
To:	John Miller Office of Nuclear Reactor Regulation
References
GL-83-11, LIC-84-038, LIC-84-38, NUDOCS 8402140092
Download: ML20080H095 (12)

Text

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Omaha Public Power District 1623 Hamey Omaha, NebrMha 68102 402/536 4000 5'

February 8, 1984 g' LIC-84-038 D Mr. James R. Miller, Chief l U. S. Nuclear Regulatory Commission i Office of Nuclear Reactor Regulation l Division of Licensing Operating Reactors Branch No. 3 Washington, D.C. 20555

Reference:

Docke t No. 50-285

Dear Mr. Miller:

Fort Calhoun Reload Core Analysis Methods and Verification Attached is additional information concerning the District's profi-ciency in using the CESEC-III computer code, as requested by members of your staff during a January 26, 1984 telephone conver-sation.

Sinceqely,

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W. C. ones Divisi n Manager

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Production Operations WCJ/JJF:jmm Attachment cc: LeBoeuf, Lamb, Leiby & MacRae 1333 New Ilampshire Avenue, N.W.

Washington, D.C. 20036 Mr. E. G. Tourigny, Project Manager Mr. L. A. Yandell, Senior Resident Inspector

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Attachment The Cycle 8 Safety Analysis Report (SER) requested that the Dis-trict submit reload methodology reports, while paying particular attention to Generic Letter 83-11 [ Reference (1)]. The District committed to providing these reports in Reference (2), based on the report scope agreed to with members of your staff in a May 11, 1983 telephone conversation.

Generic Letter 83-11 requires licensees who intend to use a safety analysis computer code to demonstrate their proficiency in using the code by submitting code verification done by themselves. As discussed in Generic Letter 83-11, verification includes compari-sons performed by the licensee of the ccde results to experimental data, plant operational data, or other benchmarked analyses.

As discussed in the District's Transient and Accident Methods and Verification report [ Reference (3)], the purpose of the District's verification is to demonstrate our proficiency in using the CESEC-III code. This proficiency is demor.strated through comparisons of CESEC-III code results with calculations performed by Combustion Engineering (CE) and Exxon Nuclear Company (ENC) and with Fort Calhoun Station transient test data.

J The District choose to compare Cycle 8 CESEC-III results with Cycle 6 CE and ENC results because the reactor physica parameters affecting the results of the Cycle 8 analyses were essentially the same as those of Cycle 6 for the events considered and no system changes were made between the cycle 6 and Cycle 8 analyses.

Although some differences do exist, they are small and their im-pact can be readily isolated. The Cycle 8 to Cycle 6 comparisons were previously presented to members of your staff and submitted in Reference (4).

To further demonstrate the validity of the Cycle 8 to Cycle 6 com-parisons, the analysis input data given in Reference (3) has been expanded and clarified, where necessary. Each of the Cycle 8 to Cycle 6 comparisons is described below.

CEA Drop As stated in Reference (3), a comparison was made between the Cycle 6 ENC analysis and the Cycle 8 OPPD analysis. Table 1 has been updated and expanded to show that:

(1) The effective moderator temperature coefficients used were

-2.76x10-4 Ap/* F and -2.7 x10-4 Ap/*F for Cycles 6 and 8, respectively.

(2) The effective doppler coefficients used were -2.556x10-5 Ap/*F and -2.496x10-5 gpf.p, (3) The' delayed neutron fractions were 0.0045 and 0.00476.

CEA Drop (Continued)

(4) The RCS flow rate, which is a third order effect, is in-creased for Cycle 8 by 7,000 gpm (approximately 3.68%).

This change is insignificant.

(5) The dropped CEA worth for Cycle 6 is 0.06% apmore. This has little effect on the final power level calculated in the analysis.

(6) All other significant parameters are identical.

Based on the above comparisons, it may be concluded that the Cycle 6 and Cycle 8 moderator temperature coefficients, doppler coeffi-cients, delayed neutron fractions, and dropped CEA worths are es-sentially the same, with all other first and second order para-meters identical. Therefore, a direct comparison may be made be-tween Cycle 6 and Cycle 8. The results of the comparison show ex-cellent agreement.

Hot Zero Power (HZP) Main Steamline Break ( MS's B )

J The HZP MSLB cases considered in Reference (3) for comparison in-clude the Cycle 6 ENC analysis, the Cycle 6 C3 analysis for the control grade automatic auxiliary feedwater system, and the Cycle 8 OPPD analysis. Table 2, which has been expanded from Reference (3), summarizes the input parameters for each of the casee. This table shows that:

(1) All cases were initiated from the same power level, with the same inlet temperature.

(2) The CE and OPPD analyses were both initiated from the maxi-mum permissible pressurizer pressure (during normal oper-ation), while the ENC analysis was initiated from the lowest permissible pressure. The difference has a negligible ef-feet on the thermal hydraulic response of the system.

(3) The CB and ENC analyses assumed the minimum RCS flow value from the PSAR. The OPPD analysis assumed the Cycle 8 Techni-cal Specification limit which had been increased from the FSAR (and original Technical Specification) minimum guaran-teed flow. This parameter has, at most, a second order ef-fect on the transient, so the flow difference is not signi-ficant.

(4) The effective moderator temperature coefficient of reacti-vity is presented in Figure 1 as a plot of reactivity versus moderator temperature. This figure shows virtually identi-cal cooldown curves for all three cases. Since this para-meter is the single most important input to the event, justi-fication for comparison of the analyses is provided by the essentially identical cooldown curves.

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a Hot Zero Power (HZP) Main Steamline Break (MSLB) (Continued)

(5) The doppler reactivity, as a function of fuel temperature, was identical for the CE and OPPD analyses which assume a bounding curve. The ENC doppler function is contained in Reference (5). Comparisons of physica data for Cycle 8 and Cycle 6 show that the doppler coefficient for. cycle 8 is essentially unchanged from Cycle 6.

(6) The doppler coefficient multipliers were both 1.15 for the CE and OPPD analyses, while a value of 0.8 was assumed for the ENC analysis.

(7) The ENC analysis assumed a shutdown margin of 3% Ap , while the OPPD analysis assumed a 4% Ap shutdown margin. The CE analysis assumed a 4.2% ao scram worth for HZP, with the most reactive CEA stuck, because the purpose of this analy-sis was to verify acceptable return-to-power results for the auxiliary feedwater system. These differences were isolated in the analysis results, as discussed in Reference (3).

(8) The initial steam generator pressures assumed for the CE and OPPD analyses were essentially the same, while the ENC as-sumed value is not available. For Cycle 8, a higher initial steam generator mass inventory was assumed. This change re-sults in a longer cooldown time (due to a longer time to steam generator dryout, particularly at HZP, as opposed to a relatively insignificant difference at HFP). This effect can be observed in Figure 4-5 of Reference (3).

(9) The delayed neutron fractions were identical for the CE and OPPD analyses, while the ENC assumed value was unavailable.

This is a second order effect.

(10) The inverse boron worth was unavailable for the ENC analy-sis, while the CE and OPPD values used were -87 and -94 pps/% Ap , respectively. This parameter is associated wit.h the ef fects of safety injection, which provides additional shutdown margin. The larger, more negative value is more conservative. This small difference has no significant effect on the analysis.

(11) The values of the gap thermal conductivity were assumed to be the minimum values. The ENC and OPPD analyses assumed the same bounding values, while the CE analysis used a conservative value derived from fuel performance calcu-lations. All three values are consistent, with no signifi-cant difference.

From the above input data comparisons, it can be observed that the CE analysis-using CESEC-I and the OPPD analysis using CESEC-III should be most nearly the came. Some differences in the results will exist which are attributable to a different scram worth and

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5 Hot Zero Power (HZP) Main Steamline Break (MSLB) (Continued) to OPPD's use of the most sophisticated code version which in-cludes modeling of wall heat, vessel upper head voiding, and the i

safe ty injection tanks, with the use of a core temperature tilt asymmetry option. With the purpose of the submittal of Section 6 of Refsrence (3) being a demonstration of the District's profi-ciency in the use of the CESEC-III computer code, as per Reference (1), the District's proficiency can be demonstrated by obtaining similar (but not necessarily identical) results for these compari-sons, in addition to benchmarking against actual plant data, as shown in Section 6.2 of Reference (3). The results of the Refer-ence (3) comparisons show excellent agreement betweer< the thermal hydraulic responses of the CE and OPPD analyses with consistent results in the reactivity feedback responses of the ENC, CE, and OPPD analyses. Insufficient data was available from the ENC analysis to make the RCS pressure and steam generator pressure ~

comparisons.

Hot Full Power (HFP) Main Steamline Break (MSLB)

The HFP MSLB cases considered in Reference (3) for comparison in-clude the same analyses as for the HZP MSLB. Table 3 shows an ex-pansion of the case input data presented in Reference (3). This j table shows that:

(1) The same comments from the HZP MSLB items (1), (3), (5),

(6), (9), and (11) apply to the HFP case.

(2) The CE and OPr0 analyses were both initiated from the maxi-mum permissible pressurizer pressure (during normal oper-ation), while the ENC analysis was initiated from the nominal pressure minus the 22 psia uncertainty. This dif-

ference has a negligible effect on the system's thermal hydraulic response.

(3) The effective moderator temperature coefficient of reacti-i vity is presented in Figure 2 as a plot of reactivity versus moderator temperature. This figure shows virtually identi-cal cooldown curves for all three cases. Since this para-meter is the single most important input to the event, justi-fication for direct comparison of the Cycle 6 and Cycle 8 cases is provided.

(4) Both the CE and ENC analyses utilized a scram worth of 5.81%

op, while a greater scram worth of 6.57% ap was available for the Cycle 8 OPPD case. The effect of this difference is.

discussed in Reference (3).

(5) The comments of item (8) of the HZP MSLB apply to steam generator input data, with the exception that the CE analy-sis used the steam generator pressure which was anticipated for Cycle 6 operation at 1500 MWt and 545*F inlet temper-ature Which was changed from the previous cycle's power and

i Hot Full Power (HFP) Main Steamline Break (MSLB) (Continued) inlet temperature of 1420 MWt and 536*F. The Cycle 8 analy-sia used the steam generator pressure that would be observed for 1530 MWt and 547'F.

(6) The inverse boron worth used in the CE and OPPD analyses varied by 11 ppm /% ap. This effect is seen in the results of safety injection which occurs after the peak return-to-power, thus having essentially no impact on the event. The input value for the ENC analysis was unavailable.

From the above input data comparisons, it can observed that the OPPD analysis more closely resembles that performed by CE. It should be again reiterated that the District employs CE method-ology and codes, so the results of these two analyses should be ccneistent. The results of the Reference (3) comparisons show excellent agreement between the CE and OPPD analyses, as antici-pated.

Conclusions J

Section 6 of Reference (3) fulfills the requirements of Reference (1) for demonstrating the District's proficiency in using the CESEC-III code for performing safety analysis in support of reload licensing. This has been done by benchmarking the code against plant transient data in Section 6.2 of Reference (3) and showing that the same (but not identical) results are obtained in compari-son of the OPPD Cycle 8 analyses of the CEA drop and MSLB to analyses performed by ENC and/or CE for Cycle 6, in which the physics and operating parameters are nearly identical. The input data assumed in the CSA drop analyses performed by OPPD and ENC are virtually identical, with the same analysis results obtained.

The OPPD MSLB analyses at HZP and HFP show excellent agreement with those performed by CE, with less but still good agreement to those performed by ENC. The variance between the ENC and CE MSLB analyses, which have been approved by the staff, is greater than the OPPD to ENC or OPPD to CE analyses, providing further as-surance of the correctness of the Cycle 8 analyses and confidence in the District's ability to correctly use the CESEC-III code.

References (1) Generic Letter 83-11, " Licensee Qualification for Performing Safety Analysis in Support of Licensing Actions",

February 8, 1983.

(2) Letter from W. C. Jones to R. A. Clark (LIC-83-184),

July 28, 1983.

(3) OPPD-NA-8303-P, " Omaha Public Power District Transient and Accident Methods and Verification", Eeptember, 1983.

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d References (Continued)

(4) CEN-242(0)-P, "OPPD Responses to NRC Questions on Port Calhoun cycle 8", February 18, 1983.

(5) XN-NF-79-79, " Fort Calhoun Cycle 6 Plant Transient Analysis Report", October, 1979.

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Table 1 COMPARISON OF PARAMETERS INCLUDING UNCERTAINTIES USED IN THE CEA DROP ANALYSES FOR CYCLES 6 AND 8 ENC OPPD Parameter Units Cycle 6 Cycle 8 Initial Core Power MWt 102% of 1500 102% of 1500 Level Core Inlet Temperature *F 547 547 Pressurizer Pressure psia 2053 2053 RCS Flow Rate gpm 190,000 197,000 Moderator Temperature 10-4 ap/*F -2.76* -2.7 Coefficient I Doppler Coefficient 10-5 ap/.F -2.13 -2.17 Doppler coefficient 1.20 1.15 Multiplier Dropped CEA Worth  % op -0.34 -0.28 Delayed Neutron 0.0045 0.00476 Fraction

• 0 PPD-NA-8303-P reported a value of -2.3x10-4 which did not in-clude the 1.20 multiplier used.

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Table 2 COMPARISON OF PARAMETERS INCLUDING UNCERTAINTIES USED IN THE HZP MSLB ANALYSIS FOR CYCLES 6 AND 8 CE ENC Cycle 6 OPPD Parameter Units Cycle 6 AFW Cycle 8 Initial Core Power MWt 0.0 1.0 1.0 Level Core Inlet Temperature *F 532 532 532 Pressurizer Pressure psia 2053 2175 2172 RCS Flow Rate gpm 190,000 190,000 197,000 Effective Moderator 10-4 Ap/*F ----------See Figure 1----------

Temperature Coefficient

, Doppler coefficient 10-5 apf.y *** **** ****

Dopplev Coefficient 0.8 1.15 1.15 Multiplier Minimum CEA Scram  % Ap -3.0 -4.2 -4.0 Worth (Shutdown Mar-gin)

Initial Steam Gener- psia

• 905** 895 ator Pressure Initial Steam Genor-  % Narrow 63 63 70 ator Mass Inventory Range Scale (Level)

Delayed Neutron

• 0.0058 0.0058 Fraction Inverse Boron Worth pgm_ *

-87 -94 (For Safety Injection)  % op H g'ap Btu _

500 868 500 hr ft' 'F

• Data used in analysis unavailable.

• • Revised from OPPD-NA-8303-P.

• • • Curve supplied in XN-NF-79-79, " Fort Calhoun Cycle 6 Reload Plant Transient Analysis Report", October, 1979. l

• • • • Same bounding doppler curve used. l l

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. o Table 3 COMPARISON OF PARAMETERS INCLUDING UNCERTAINTIES USED IN THE ilFP MSLB ANALYSES FOR CYCLES 6 AND 8 CE ENC Cycle 6 OPPD Parameter Units Cycle 6 AFW Cycle 8 Initial Core Power MWT 102% of 102% of 102% of Level 1500 1500 1500 Core Inlet Temperature *F 547 547 547 Pressurizer Pressure paia 2078 2175 2172 RCS Flow Rate gpm 190,000 190,000 197,000 Moderator Temperature 10-4 Ap/

• F ----------See Figure 2----------

Coefficient Doppler Coefficient 10-5 apj. y *** **** ****

DoppleefCoefficient 0.8 1.15 1.15 Multiplier Minimum CEA Scram  % ap -5.81 -5.81 -6.57 Worth Initial Steam Gener- psia

• 858** 890 ator Pressure Initial Steam Gener-  % Narrow 63 63 70 ator Mass Inventory Range Scale (Level)

Delayed Neutron

• 0.0058 0.0058 Fraction Inverse Boron Worth *

-87 -98 li gap Btu 500 568 500 hr ft4 'F Data used in analysis unavailable.

• • Revised from OPPD-NA-8303-P.

• • • Curve supplied in XN-NF-79-79, "Fcrt Calhoun Cycle 6 Reload Plant Transient Analysis Report", October, 1979.

• • • • Same bounding doppler curve used.

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200 300 400 500 600 700 CORE AVERAGE MODERATOR TEMPERATURE,*F NOTE: CYCLE 6: ENC ANALYSIS CYCLE 6 AFW: CE ANALYSIS CYCLE 8: OPPD ANALYSIS l

l Stem Line Break Incident (HZP) OmahaPublicPowerDistrict figure Reactivityvs.HoderatorTemperature FortCalhounStation-UnitNo.1 i

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l SteasLineBreakIncidentOfP; OmahaPublicPowerDistrict figure lleactivity vs. bderator Tesperature Fort Calhoun Station-thit b. I 2 L

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