ML19296B436

Submits Supplemental Matl Supporting 790717 Application to Amend OL DPR-40 Increasing Power Level.Forwards Small Break LOCA Evaluation,Revised Tech Specs & Performance Evaluation for High Burnup Demonstration Assembly

ML19296B436
Person / Time
Site:	Fort Calhoun
Issue date:	02/12/1980
From:	William Jones OMAHA PUBLIC POWER DISTRICT
To:	Reid R Office of Nuclear Reactor Regulation
Shared Package
ML19296B432	List: ML19296B430 ML19296B436 ML19296B439 ML19296B448
References
NUDOCS 8002200560
Download: ML19296B436 (34)

Text

d$ ,a Omaha Public Power District 1623 HARNEY a OMAHA, NEBRASKA 68102 a TELEPHONE 536 4000 ARE A CODE 402 February 12, 1980 Director of Nuclear Reactor Regulation ATTN: Mr. Robert W. Reid, Chief Operating Reactors Branch No. 4 U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Reference:

Docket No. 50-285 Gentlemen:

Omaha Public Power District hereby submits supplemental material in support of (1) the Application for Amendment of Operating License

(" Stretch Application"), filed July 17, 1979, which seeks to amend Facility Operating License No. DPR-40 to permit Cycle 6 operation fol-lowing core reload at an increased power level of 1500 MWt, and (2) the Application for Amendment of Operating License (" Reload Application"),

filed July 17, 1979, which seeks to permit Cycle 6 operation follow-ing core reload. Copies of the following materials are enclosed:

(1) " Fort Calhoun Cycle-6 Small Break LOCA Evaluation at 1420 MWT" (40 copies).

(2) Responses to NRC Questions 22 and 23 on " Fort Calhoun Cycle 6 Reload Plant Transient Analysis Report, XN-NF-79-79" (40 copies).

(3) " Fort Calhoun Unit 1 Cycle 6 Performance Evaluation for High Burnup Demonstration Assembly" (40 copies).

(4) Revision to proposed Technical Specifications, which includes TM/LP equation (40 copies).

(5) Discussion supporting item (4), revised Technical Specifications (40 copies).

(6) " Revised Limiting Large Break LOCA Analyses for Fort Calhoun Using the ENC WREM-11A PWR ECCS Evaluation Model" (20 copies -

20 additional copies will be submitted under separate cover).

Should you desire additional information on these materials, please advise us.

3;ih 6 ely, j

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C.' Jones Divis' ion Manager Production Operations WCJ/KJM/BJH:jmm Enclosures cc: LeBoeuf, Lamb, Leiby & MacRae ,,

1333 New Hampshire Avenue, N. W. 8002200 Washington, D. C. 20036

9 4

Fort Calhoun Cycle-6 Small Break LOCA Evaluation at 1420 MWT

l l

1.0 IttTRODUCTIOil AfiD

SUMMARY

The small break LOCA ECCS performance evaluation for Ft. Calhoun presented herein, demonstrates conformance with 10CFR50.46 which contains the Acceptance i

Criteria for Emergency Core Cooling Systems for Light-Water-Cooled ReactorsO) .

j This evaluation demonstrates acceptable ECCS, performance for Ft. Calhoun at a i

core power level of 1443 (102% of 1420 Mwt) and a peak linear heat generation j rate (PLHGR) of 15.5 kw/f t. The method of evaluation and the results are presented in the following sections.

2.0 METHOD OF A!!ALYSIS I

I The analysis consisted of a comparison of, the pertinent parameters for Ft.

Calhoun Cycle 6 with those employed in, the most recent small break LOCA analysis (Reference 2) performed for Cycle 3 at a power level of 1448 Mwt (102% of 1420 Mwt) and a PLHGR of 15.5 kw/ft. A 1imited number of parameter differences were identified and their impact is discussed in the report.

3.0 RESULTS ,

i A comparison of all pertinent system parameters for Ft. Calhoun Cycle 3 and Cycle 6 was completed. Table 1 lists the major parameters for each cycle along with any parameter differences resulting from the introduction of EXXON assemblies into Cycle 6. The comparison demonstrates that the Reference 2 small break LOCA analysis results remain conservative and applicable to Cycle 6..

The reasons supporting this conclusion are enumerated below:

1. The Cycle 6 core contains a mixture of CE and EXXON fuel assemblies. The pressure drop across the core for Cycle 6 is 7.51 psia (core composed of EXX0ff and CE fuel) compared to the Reference 2 analysis value of 7.49 psia

........,f,. i i .c vuiue us use Lure pressure cr0p docs not significantly influence the small break response since quiescent pool boiling is achieved prior to the potential for core uncovery. Thus, this very small change in Cycle 6 will have no effect on the predicted results.

2.

The EXXOtt fuel average tenperature for Cycle 6 is conservatively calculated to be no greater than 175 F higher than the Reference 2 analysis value for the average fuel temperature for the CE fuel. Also, the'EXXOtt inter-nal fuel pin pressure is conservatively calculated to be no greater than 125 psi higher than the Reference 2 analysis value for internal pin pressure for the CE fuel. The small break LOCA peak clad temperature is insensitive to fuel characteristics (fuel average temperature and fuel internal pin pressure) since during the period of interest following a small break LOCA (the period of time during which the core may partially uacover), the peak clad temperature is dependent on the core decay heat generation rate. That is, prior to the potential uncovery time during the LOCA, all of the initial fuel stored energy will be removed from the core rendering the.small break LOCA ECCS performance insensitive to initial fuel stored energy. The decay heat level during the period of uncovery will be proportional to the initial PLHGR only, 15.5 kw/ft for both the Cycle.6 and the Refer-ence 2 analyses.

The higher pin pressure also has no impact since the peak clad temperature is not sufficient to cause fuel pin rupture.

In fact, for the limiting small break (that which does not result in safety injection tank actuation' the 0.075 ft break of Reference 2) an increase in pin pressure of at least.320 psi is required to cause rupture at the PCT of 1593 F for this break. Therefore, the increase in initial pin pressure for Cycle 6 of 125 psi will not cause rupture.

3.

The Cycle 6 core inlet and outlet temperatures, 547 F and 601.4 F, respectively, are higher than the Cycle .3 values of 540 F and 593 F.

respectively.

These nominally higher temperatures will not adversely affect ECCS performance for the following reasons: the higher initial temperature will not delay either reactor trip or SIAS since the saturation pressure corresponding to the maximum outlet temperature is below the minimum low pressurizer pressure trip setpoint of 1728 psia and the minimum SIAS setpoint of 1578 psia.

These setpoints will still be achieved during the subcooled decompression period which is controlled by the assumed break size.

Also, later during the small break LOCA, the RCS pressure achieves a

pressure plateau which is controlled by the steam generator secondaries and is independent of initial coolant temperatures. Therefore, the increases

in RCS coolant temperatures will have no effect on RCS pressure for the duration of the transient.

Since'the differences between the pertinent system parameters for Cycle 6 and Cycle 3 are minimal, the Reference 2 analysis is applicable to Cycle '6.

4.0 CONCLUSION

The small break LOCA evaluation for Ft. Calhoun Cycle 6 demonstrates conformance with the acceptance criteria of 10CFR50.46. The results also demonstrate a significant margin below the acceptance limits, reaffirming that the large break LOCA would yield more limiting results, and would therefore define the maximum allowable PLHGR.

5.0 REFEREitCES 1.

Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Cooled Nuclear Power Reactors, Federal Register, Vol. 39, No. 3 -

Friday, January 4, 1974.

2. Cycle 3 Small Break LOCA Analysis (Letter from OPPD to NRC to be supplied by OPPD).

• Y e

TABLE 1 General System Parameters Quantity Cycle 3 Cycle 6 Reactor power level (102% of flominal) 1448 1448 MWt Average linear heat rate (102% of flominal) 5.85 5.85 kw/ft Fuel centerline temperature at peak linear 3945.8 <4120.8

• F heat rate Hot rod gas pressure 1346.6 <l 4 71. 6 *

• psia Peak linear heat rate 15.5 15.5 kw/ft Reactor vessel inlet temperature 540 547 F Reactor vessel outlet temperature 593 601.4 F Active core height 10.7 10.7 ft Total core pressure drop 7.49 7.51 psi

• EXXOil fuel only. The EXX0fl fuel temperature is conservatively calculated to i

be no more than 175 F higher than the CE fuel at 15.5 kw/f t.

• • EXXOfi fuel only. The EXXOtt fuel pin internal pressure is conservatively calculated to have a pin pressure of no more than 125 psi higher than the CE fuel at 15.5 kw/f t.

I i

9 I

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s Responses to NRC Questions 22 and 23 on " Fort Calhoun Cycle 6 Relcad Plant Transient Analysis Report, XN-NF-79-79"

~ -

Question 22 It is our understanding that the TM/LP equation to be used is not the equation given on Page 30 (XN-NF-79-77). Justify the validity of the analysis presented in References 1 and 2 in view of the changed equation.

Response '

The impact of the final TM/LP on the analyses reported in XN-NF-79-79 has been carefully considered. The transients listed in Category A of Table 1.1 (XN-NF-79-79) which were reported to have tripped on TM/LP have been reanalyzed using the revised TM/LP.

Results of the reanalysis are reported in the discussion of the Application for Amendment of Facility Operating License forwarded herewith. Those results indicate that modification of the RPS by inclusion of the revised TM/LP does not alter the protection afforded by the RPS.

Question 23 Describe how delays in the TM/LP trip circuitry (including the initial delays in power and inlet temperature) are modeled in PTSPWR for the Fort Calhoun safety analyses.

Response

The overall scram delay time reported in XN-NF-79-79 for the low pressurizer pressure subsumes signal acquisition and processing delays inherent in the TM/LP trip circuitry. The overall scram time for the low pressurizer pressure used in Fort Calhoun safety analysis PTSPWR calculations is consistent with the Fort Calhoun FSAR.

Fort Calhoun Unit 1 Cycle 6 Performance Evaluation for High Burnup Demonstration Assembly e

TABLE OF C0flTEllTS Table of Contents List of Figures List of Tables

1. Introduction and Summary

2. Program Description

3. Assembly Performance

4. Environmental Assessment

5. ECCS Performance

6. References

List of Figures Figure Number Title 1 Clad Temperature of Neighboring Pins 2 Peak Clad Temperature 3 Hot Spot Gap Conductance 4 Peak Local Clad 0xidation 5 Clad Temperature, Centerline Fuel Temperature, Average Fuel Temperature and Coolant Temperature for the Hottest Node 6 Hot Spot Heat Transfer Coefficient 7 Hot Rod Internal Gas Pressure

List of Tables Table ilumber Title 1 Summary of Results for Ft. Calhoun Cycle 6 Batch D ECCS Perfo mance 2 Variables Plotted as a Function of Time 3 General System Parameters

1. Introduction and Summary 1.1 Introduction This document provides a performance evaluation for the proposed exposure of a high burnup demonstration assembly in Cycle 6 of the Fort Calhoun reac tor. This demonstration is part of the improved fuel utilization program funded by the Department of Energy. It involves a fif th cycle exposure for a Batch D assembly originally introduced in Cycle 2, resulting in a discharge burnup as high as 45,000 MWD /MTV. A summary of the program and the demonstration is given in Section 2 of this report.

1.2 Summary This report addresses the impact of high burnup on the performance of the demonstration assembly. Section 3 gives the conclusion of assembly performance evaluations including clad collapse and fuel rod bow An environmental assessment is included as Section 4.

A review of all postulated accidents and anticipated operational occurences addressed in Cycle 5 (Reference 1) has shown that only the ECCS evaluation would be impacted by the high burnup of the demonstration assembly. The assembly was included in the design and performance evaluations for Cycle 6 (References 2, 3, 4). How-ever, the impact of the high burnup itself was not addressed. There-fore, an ECCS evaluation is included as Section 5.

The performance evaluation for the demonstration assembly is based on a Cycle 5 length of 10,500 MWD /MTV and a projected Cycle 6 length of 10,000 MWD /MTU. It is applicable to any combination of cycle lengths no greater than a two-cycle length of 20,500 MWD /MTU.

This evaluation is applicable to the operating conditions and Technical Specifications of Cycle 5 (Reference 1) or the proposed conditions and Technical Specifications of Cycle 6 (Reference 2) including the increase in power level to 1500 MWt.

2. Fuel Utilization Program Description The high burnup demonstration assembly is part of a Department of Energy program to improve fuel utilization through more efficient fuel management and an increase in fuel burnup.

More efficient fuel management will be achieved throu' h the im-plementation of a low leakage concept called SAVFUEL (Shinmed And Very Flexible Uranium Element Loading), which is expected to reduce uranium requirements' b~y two to four percent (2% to 4%). In addition, the burnup will be increased sufficiently to reduce uranium requirements by five percent (5%) for the reactor's intended future operating made (eighteen-month cycle) relative to the current operating mode (yearly cycle).

The proposed program will be accomplished in two phases. Phase I consists of two leaa bundle demonstrations. Use of SAVFUEL fuel management causes some of the fuel rods to experience a different power history from cycle to cycle in comparison to more conventional fuel management. Therefore, Phase IA will include a demonstration that this duty cycle does not have a deleterious effect on fuel performance. This demonstration involves four Batch G assemblies -

first inserted in Fort Calhoun in Cycle 5. These assemblies will experience a high power-low power-high power duty cycle during Cycles 5, 6 and 7, respectively,in order to demonstrate acceptable fuel performance. A performance evaluation will be performed prior to their insertion into a high power region for Cycle 7.

Phase IB involves the high burnup demonstration D assembly to be irradiated for its fifth cycle in Cycle 6. Poolside and hot cell examinations are planned for this assembly before and after Cycle 6 and for other D assemblies which are discharged after Cycle 5. In addition, examinations are planned for Batch C assemblies discharged af ter Cycle 4. Included in the examination will be measurements of internal fuel rod gas pressure, dimensional changes and clad corrosion, and inspection of pellet and clad microstructure.

Batch C fuel discharged after Cycle 4 received rod average exposures up to 39,000 MWD /MTU while Batch D fuel will have received up to 40,000 MWD /MTU by the end of Cycle 5.

Related experience with high burnup for C-E fuel includes EPRI test rods in Calvert Cliffs Unit I receiving burnups as high as 40,000 MWD /MTU at the end of Cycle 3 (April 1979) and planned exposure to 47,000 MWD /MTU through Cycle 4 (discharge mid-1980).

Pending successful completion of Phase I, a large scale Phase II demonstration is anticipated in which SAVFUEL fuel management will be used core-wide, and batch burnups will be increased by increasing the cycle length without increasing the reload batch size. Implementation of Phase II will reduce uranium requirements by seven to nine percent (7% to 9%).

3. Assembly Performance 3.1 Mechanical Design The mechanical design of the high burnup demonstration assembly is described in detail in Reference 5.

C-E has performed an analytical prediction of clad,iing creep-collapse time for the demonstration assembly. Using the computer code CEPAN (Reference 6), C-E concludes that no creep-collapse will be experienced by this assembly during Cycle 6.

Time to cladding creep-collapse for the demonstration assembly is pre-dicted to be greater than 45,000 EFPH while the cumulative exposure expected at the end of Cycle 6 is less than 40,000 EFPH.

3.2 Fuel Rod Bowing Effects Fuel rod bowing effects on DNB margin for the high burnup demonstration assembly during Cycle 6 have been evaluated with the guidelines set forth in Reference 7. Since the demonstration assembly reaches a burnup of less than 45,000 MWD /T at end of Cycle 6, the fuel rod bowing penalty on DNB prescribed by Reference 7 would be less than 7%. How-ever, the assembly never achieves radial peaks within 30% of the max-imum radial peak in the core at any time during Cycle 6. Therefore, no penalty on core DNB margin is required.

Fuel rod bow 1ng effects on DNB margin are now being incorporated in safety and setpoint analyses of C-E designed plants and reloads in the manner described in Reference 8. This reference contains penalties on minimum DUBR due to fuel rod bowing as a function of burnup generated using NRC guidelines contained in Reference 9. The penalty associated with burnups up to 50,000 MWD /T would be less than 1% in this case.

Therefore, the 30% margin in radial peaking factor for the demonstration assembly is considerably greater than the penalty required to account for fuel rod bowing.

3.3 Fuel Performance Evaluations Fuel performance evaluations on the denonstration assembly, including gap conductance, fuel tenperature and effects of densification, were performed using the model described in Reference 10.

4. Environmental Assessment Examination of potential environmental impact of the irradiation of one demonstration D assembly for a fifth exposure cycle yields the conclusion that the impact would be negligible. The demon-stration increases burnup to 49,000 ffdD/MTU (rod average) in one assembly, larger than the maximum demonstrated burnup of current C-E operating reactors (40,000 MWD /MTU rod average). This increased burnup leads to a larger inventory of long-lived fission products in the fuel.

However, the impact of this demonstration is judged to be small for the following reasons: 1) small number of fuel rods experiencing high burnup; 2) fuel examinations of this assembly and other similar assemblies during refueling to insure fuel rod integrity; 3) lower fission product inventory in other assemblies due to reduction in the core wide available excess reactivity; 4) placement of the assembly in a low power location; and 5) continued compliance with Technical Specifications on reactor coolant radioactivity (2.1.4) to maintain concentrations of radioactivity within the allowed limits.

In addition, if the data to be obtained from the demonstration assembly permits future full core implementation of higher burnups, there would be a future net reduction in environmental impact due to the concomitant reduction in uranium mining, milling, separation work, fuel fabrication, and spent fuel storage.

5. ECCS Performance 5.1 Introduction and Sunmary The ECCS performance evaluation for Ft. Calhoun Cycle 6 Batch D fuel presented herein demonstrates appropriate conformance with the Acceptance Griteria for Light-Water-Cooled Reactors as presented in 10CFR50.46. l ll ) This evaluation only applies to the single Batch D fuel a s sembly. The results of this analysis indicate acceptable ECCS performance for the Batch D assembly at a linear heat generation rate (LHGR) of 10.0 kw/ft. The method of analysis and results are presented in the following sections.

5.2 Method of Analysis The Cycle 6 Batch D fuel consists of one fuel assembly located at the center of the reactor core. If the maximum allowable peak LHGR in the core is limited to 15.22 kw/ft for Cycle 6, the highest power pin in Batch 0 has an expected peak LHGR of less than 10.0 kw/ft. The method of analysis employed demonstrates that the low linear heat generation in the hottest Batch D fuel rod results in a much less limiting response than the higher power assemblies in Cycle 6.

Burnup dependent calculations were performed using the FATES (12) and STRINKIN-II(13) computer codes to determine the limiting condition for the ECCS performance analysis of the Batch D assembly. The break size and type analyzed,1.0 DEG/PD*, js the same as was analyzed in the previous limiting break analysistl4). The radiation enclosure in the Cycle 6 STRIKIN-II analysis used a conservative adaptation of the peak clad temperature versus time profile (Figure 1) for the sur-rounding pins in the higher power neighboring assemblies, which represents their radiation heat transfer contribution to the hot pin in the Batch D assembly. The more limiting pins in the Batch E and G assemblies adjacent to the D the CE radiation heat transfer model(aspembly were analyzed 141 as approved using by NRC. A variation was necessary to conservatively represent the thermal response of those high power pins in the neighboring assemblies which are adjacent to the less limiting D pin.

5.3 Results Table 1 presents the analysis results for the limiting 1.0 DEG/PD break. A list of the significant parameters displayed graphically is presented in Table 2. A summary of the fuel and system parameters is shown in Table 3.

It should be noted that this analysis was performed at a power level of 1448 MWt (102% of 1420 t'Wt). Since the response of the highest power Batch D pin is much less limiting than the 10CFR50.46 acceptance limits, it is clear that the conclusion derived from this analysis is equally valid at the 5.6% higher power.

• l.0 x Double-Ended Guillotine Break in the Reactor Coolant Pump Discharge leg

5.4 Conclusion As can be seen from the results, the worst break analysis yields a peak clad temperature of 1303 F which is well below the cri-teriatll) limit. Since the maximum clad tenperatures achieved are low enough, pin rupture is not predicted and the maximum local zircor,ium oxide percentage is only 0.13%. These results were derived for a power level of 1448 MWt. In view of the large margins, it is expected that the same conclusions would be demonstrated at the power level of 1530 MWt.

5.5 Computer Code Version Identification The NRC approved version 77063 of the STRIKIN-II code of Combustion Engineering's ECCS Evaluation Model was used to perform the burnup dependent calculations in evaluating the fuel data and in predicting the clad temperature response shown herein.

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TABLE 1 Summary of Results for Ft. Calhoun Cycle 6 Batch D -ECCS Perfonnance Results Break: 1.0 DEG/PD Peak Clad Temperature: 1303 F Time of Peak Clad Temperature: 10 sec Maximum Local Zirconium -

0xide(%): .13

TABLE 2 Ft. Calhoun Cycle 6 Batch D Variables Plotted as a Function of Time Variables Figure Designation Peak Clad Temperature 2 Hot Spot Gap Conductance 3 Peak local Clad Oxidation 4 Clad Temperature, Centerline Fuel Temperature, Average Fuel Tempe ature and Coolant Temperature for Hottest flode 5 Hot Spot Heat Transfer Coefficient 6 Hot Rod Internal Gas Pressure 7 o

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4

TABLE 3 Ft. Calhoun Cycle 6 Batch D General System Parameters Quantity Value Reactor Power Level (102% of Nominal) 1448 MWt Linear Heat Generation Rate (LHGR) for the Batch D fuel 10.0 kw/ft Gap Conductance at LUGR 2000 BTU /hr ft2 op Fuel Centerline Temperature at LHGR 2200 *F Fuel Average Temperature at LHGR 1451 F Hot Rod Gas Pressure 1581 psia Hot Rod Burnup 49,963 MWD /MTU

FIGURE 1 FT, Call 10Ull CYCLE 6 BATCil D 1.0 x DOUBLE EI1DED GUILLOTlfiE BREAK AT PUHP DISCl1ARGE LEG 2200 -

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FIGURE 6 FT. CALil0UN CYCLE 6 BATCil D

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6. References

1. Fort Calhoun Unit #1 Cycle 5 Core Reload Application, Docket No. 50-285, dated August 1978.

2. EXXON Nuclear Company, Inc., " Fort Calhoun Nuclear Plant Cycle 6 Safety Analysis Report," XN-NF-79-77, October 1979.

3. EXXON Nuclear Company, Inc., " Fort Calhoun Cycle 6 Reload Plant Transient Analysis Report," XN-NF-79-79, October 1979.

4. EXXON Nuclear Company, Inc., " Fort Calhoun LOCA Analyses at 1500 MWT Using ENC WREM-11A PWR ECCS Evaluation Model,"

XN-NF-79-89, September 1979.

5. Fort Calhoun Unit #1 Cycle 2 Core Reload Application, Docket No. 50-285, dated March 1975.

6. CEPAN Topical Report, CENPD-187, May 29,1975.

7. " Interim Safety Evaluation Report on Effects of Fuel Rod Bowing on Thennal Margin Calculations for Light Water Reactors,"

NRC Report.

8. Supplement 3-P (Proprietary) to CENPD-225P, " Fuel and Poison Rod Bowing," June 1979.

9. Letter - D. B. Vassallo (NRC) to A. E. Scherer (C-E), June 12, 1978.

10. "C-E Fuel Evaluation Model Topical Report," CENPD-139, July 1,1974.

11. Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Cooled Nuclear Power Reactors, Federal Register, Vol. 39, No. 3 - Friday, January 4, 1974.

12. CEMPD-139, "C-E Fuel Evaluation Model," July 1974 (Proprietary).

13. CENPD-135, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," April 1974 (proprietary).

CENPD-135, Supplement 2, "STRIKIN-II, A Cyclindrical Geometry Fuel Rod Heat Transfer Program (Modification)" February 1975 (Proprietary).

CENPD-135, Supplement 4, "STRIKIN-II, A Cylindrical Geometry Fuel rod Heat Transfer Program," August 1976 (Proprietary).

CENPD-135, Supplement 5, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," April 1977 (Proprietary).

14. Fort Calhoun Unit #1 Cycle 4 Core Reload Application, Docket No. 50-285, dated July 1977.