ML20126J181

Cycle 7 Final Reload Safety Analysis

ML20126J181
Person / Time
Site:	Millstone
Issue date:	06/30/1985
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20126J126	List: B11561, Forwards Reload Safety Analysis in Support of Cycle 7 Reload.Core Reload Will Not Adversely Affect Plant Safety ML20126J181
References
TAC-56814, NUDOCS 8506100524
Download: ML20126J181 (22)

Text

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Docket No. 50-336 i

Attachment Millstone Nuclear Power Station, Unit No. 2 Cycle 7 Final Reload Safety Analysis i

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TABLE OF CONTENTS Section Title Page

1.0 INTRODUCTION

AND

SUMMARY

1 1.1 OBJECTIVES 1 1.2 GENERAL DESCRIPTION 1

1.3 CONCLUSION

S 3 2.0 MECHANICAL DESIGN 4 2.1 GENERAL DISCUSSION 4 3.0 THERMAL AND HYDRAULIC DESIGN 5 4.0 NUCLEAR DESIGN 6 5.0 ACCIDENT ANALYSIS 7

5.1 INTRODUCTION

AND

SUMMARY

7 5.2 ACCIDENT EVALUATION 7 5.2.1 KINETICS PARAMETERS 8 5.2.2 SHUTDOWN MARGIN 8 5.2.3 CEA WORTHS' 8 5.2.4 CORE PEAKING FACTORS 8 5.3 INCIDENTS REANALYZED 8 5.3.1 COMPLETE LOSS OF REACTOR COOLANT FLOW 8

6.0 REFERENCES

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LIST OF TABLES Table Title Page 1 Millstone Unit 2 Cycle 7 Core Loading 11 2 Millstone Unit 2 Kinetics Characteristics 12 3 Shutdown Requirements and Margins 13 4 Sequence of Events Loss of Coolant Flow 14 me-ames jj

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LIST OF FIGURES Figure Title Pm 1 Core Loading Pattern 15 2 Zoned-enrichment Fuel. Assembly Lattice 16 3- Millstone 2 - Safety Analysis Loss of Flow - 17 Core Flow Versus Time 4 Millstone 2 - Safety Analysis Loss of Flow - 18 Nuclear Power and Heat Flux Versus Time 5 Millstone 2 - Safety Analysis Loss of Flow - 19 DNB Ratio Versus Time mnemnu j$$

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1.0 INTRODUCTION

AND

SUMMARY

l 1.1 OBJECTIVES

. l This report presents an evaluation for Millstone Nuclear Power Station Unit 2, Cycle 7, which demonstrates that the core reload will not adversely affect the safety of the plant. This evaluation was accomplished utilizing the methodology described in Reference 1.

Based upon the above referenced methodology, only those incidents analyzed and reported in the Basic Safety Report (2) (BSR) which could potentially be affected by fuel reload have been reviewed for the Cycle 7 design described herein. The results of new analyses are included and the justification for the applicability of previous results for the remaining incidents is provided.

1.2 GENERAL DESCRIPTION j

The Millstone II reactor core is comprised of 217 fuel assemblies arranged in the configuration shown in Figure 1. Each fuel assembly has a skeletal structure consisting of five (5) zircaloy guide thimble tubes, nine (9) grids (eleven assemblies have zircaloy grids, two hundred six assemblies have Inconel grids), a stainless steel bottom nozzle, and a stainless steel top nozzle. One hundred seventy-six fuel rods are arranged in the grids to form a 14x14 array. The fuel rods consist of slightly enriched uranium dioxide ceramic pellets contained in Zircaloy-4 tubing which is plugged and seal welded at the ends to encapsulate the fuel.

Nominal core design parameters utilized for Cycle 7 are as follows:

Core Power (Mwt) 2700 System Pressure (psia) 2250 Reactor Coolant Flow (GPM) 350,000*

Core Inlet Temperature (*F) 549 Average Linear Power Density (kw/ft) 6.067 (based on best estimate hot, densified core average stack height of 136.4 inches)

• Minimum guaranteed safety analysis flow mm. e2 s 1

v The core loading pattern for Cycle 7 is shown in Figure 1. The feed fuel for the Millstone II, Cycle 7 core will consist of twenty-four (24) zoned enrichment interior feed assemblies, each containing sixty (60) fuel rods at 2.62 w/o U235 and one-hundred sixteen (116) fuel rods at 2.91 w/o U235, and forty eight (48) zoned-enrichment peripheral assemblies, each containing sixty (60) fuel rods at 2.91 w/o U235 and one-hundred sixteen (116) fuel rods at 3.29 w/o U235. The zoned-enrichment assembly configuration is shown in Figure 2. The feed fuel will replace twenty (20) Combustion Engineering (CE) Batch A assemblies, one (1) CE Batch B assembly, and fifty-one (51) Westinghouse Batch F assemblies. An additional five (5)

Westinghouse Batch F assemblies will be discharged from the end of Cycle 6, and will be replaced by five (5) Westinghouse Batch F assemblies which were removed from the core at the end of Cycle 5. Due to fuel defects in Cycle 6 and subsequent symmetry considerations, fourteen (14) Westinghouse Batch G assemblies, seven (7) Westinghouse Batch F assemblies (these Batch F and G assemblies were removed from the core at the end of Cycle 5), and four (4) CE Batch A assemblies (discharged at the end of Cycle 1) are needed as well. As as result of fuel reconstitution, the fuel rods from seven (7) Westinghouse reload assablies to be used in Cycle 7 have been placed in Combustion Engineering (CE) skeletons. Also, twenty-one (21) fuel rods have been replaced with stainless steel rods in Cycle 7. The twenty-one stainless steel rods are distributed among eleven (11) fuel assemblies, with the number of stainless steel rods in each of these assemblies ranging from one to five.

A summary of the Cycle 7 fuel inventory is given in Table 1.

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2.0 MECHANICAL DESIGN 2.1 -GENERAL DISCUSSION The mechanical design of the Cycle 7 fuel assemblies is essentially identical to that of the Cycle 6 assemblies (7) The Westinghouse fuel assemblies are designed to be fully compatible with all resident Millstone 2 fuel assemblies and core components (e.g. adequate clearances for insertion of CEA's, plugging devices,etc.). Table 1 summarizes pertinent design parameters of the various fuel regions.

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3.0 THERMAL AND HYDRAULIC DESIGN A description of the thermal and hydraulic design of the Westingnouse Millstone 11 reload fuel assembly to be utilized in Cycle 7 is given in Chapter 3 of the BSR.

As discussed in the BSR, the Westinghouse fuel assemblies have been designed and shown through testing to be hydraulically compatible with all resident Millstone II fuel assemblies. The stainless steel rods in the reconstituted fuel assemblies were treated as heated rods in the THINC DNB analysis. This is conservative since it results in higher subchannel enthalpy predictions.

No significant variations in thermal margins result from the Cycle 7 reload.

The Cycle 7 analysis takes a partial credit of 3.0% of the net conservatism which exists between convoluting and summing the uncertainties of various measured plant parameters in power space. This partial credit was applied in previous cycles and is discussed in n. ore detail in the Cycle 4 Reload Safety Evaluation Report ( 5) ,

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4.0 NUCLEAR DESIGN The Westinghouse nuclear design procedures, computer programs, and calculation

, models utilized in the Millstone II, Cycle 7 reload design are presented in the BSR. Similar to the Cycle 6 evaluation (7) , Cycle 7 accident simulations take credit for the variable high power trip by terminating accidents 5% above the variable high power trip. Also PL values (see BSR Section 6.0) are computed only if the maximum allowed power density of 21 kw/ft is exceeded.

The Cycle 7 core loading results in a maximum linear heat rate of less than 15.6 kw/ft at all fuel heights at rated power. The safety analysis has specifically included the 21 stainless steel rods. Table 2 provides a summary of changes in the Cycle 7 kinetics characteristics compared with the current limit based on the reference safety analysis.(2,5,6,7) It can be seen from the table that all of the Cycle 7 values fall within current limits with a small exception in the delayed neutron fraction noted in Table 2. Table 3 provides the contol rod worths and requirements at the most limiting condition during the cycle. The required shutdown margin is based on accident analyses presented in Section 5.0.

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• l 5.0 ACCIDENT ANALYSIS '

5.1 INTRODUCTION

AND

SUMMARY

The power capability of Millstone II is evaluated considering the consequences of those incidents examined in the BSR(2) , using the previously accepted design basis specified in Section 1.2. It is concluded that the core reload will not adversely affect the ability to safely operate at 100% of rated power during Cycle 7. For the overpower transient, the fuel centerline temperature limit of 4700*F can be accommodated with margin in the Cycle 7 core. The burnup dependent densification model(3,4) was used for fuel temperature evaluations. The LOCA limit at rated power can be met by maintaining peak linear heat rates at or below 15.6 kw/ft.

5.2 ACCIDENT EVALUATION The effects of the reload on the design basis and postulated incidents analyzed in the BSR(2) and updated in the Cycle 4, 5, 6, and preliminary Cycle 7 reload safety analyses (5,6,7,8) were examined. In most cases, it was found that the effects were accommodated within the conservatism of the initial assumptions used in the previous safety analysis. For those incidents which were reanalyzed, it was determined that the applicable design bases are not exceeded, and, therefore, the conclusions presented previously are still valid.

A core reload can typically affect accident analysis input parameters in the following areas: core kinetic characteristics, shutdown margin, CEA worths, and core peaking factors. Cycle 7 parameters in each of these areas were examined as discussed below to ascertain whether new accident analyses were required.

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p-5.2.1 KINETICS PARAMETERS A comparison of Cycle 7 kinetics parameters with the current limits, established by the BSR and Cycles 4, 5, and 6 reload safety analyses, is presented in Table 2. The maximum delayed neutron fraction and the least negative doppler power coefficient above 30% power exceed the current limit

.slightly. Theseparameterswereevaluated,(7)andthepreviousanalysis was determined to be adequate. Therefore, no reanalysis is required.

5.2.2 SHUTDOWN MARGIN Changes in minimum shutdown margin requirements may impact the safety analyses, particularly the steamline break and boron dilution accident analyses. Cycle 7 shutdown margin requirements are the same as Cycle 6.

5.2.3 CEA WORTHS Changes in CEA worths may affect shutdown margin. Table 3 shows that the Cycle 7 shutdown margin requirements are satisfied.

5.2.4 CORE PEAKING FACTORS All core peaking factors for Cycle 7 were within the reference cycle limits.

5.3 INCIDENTS REANALYZED 5.3.1 COMPLETE LOSS OF REACTOR COOLANT FLOW The loss of flow accident was reanalyzed for Cycle 7 assuming a +0.4x10 -4 ap/*F moderator temperature coefficient as specified in the Millstone II Technical Specifications and Table 2, and utilizing cycle specific design parameters.

Table 4 gives the time sequence of events for this accident. The reactor coolant flow, nuclear power, heat flux, and DNB transients are shown in Figures 3, 4 and 5.

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The results show*that the reactor coolant pump speed sensing system provides sufficient protection against clad and fuel damage. The calculated minimum DN8R is 1.30.

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6.0 REFERENCES

1. Bordelon, F. M., et. al., " Westinghouse Reload Safety Evaluation Methodology", WCAP-9273, March, 1978.

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2. Millstone Unit 2, " Millstone Unit 2 Basic Safety Report", Docket No.

50-336, March, 1980.

3. Miller, J. V. (Ed), " Improved Analytical Model used in Westinghouse Fuel Rod Design Computations", WCAP-8785, October, 1976.

4. Hellman, J. M. (Ed.), " Fuel Densification Experimental Results and Model for Reactor Operation", WCAP-8219-A, March 1975.

5. Letter, Counsil to Clark, Millstone Nuclear Power Station Unit No. 2, Cycle 4 Refueling - Reload Safety Analysis, June 3,1980

6. Letter, Counsil to Clark, Millstone Nuclear Power Station Unit No. 2, Cycle 5 Refueling - Reload Safety Analysis,' November 17, 1981.

7. Letter, Counsil to Miller, Millstone Nuclear Power Station Unit No. 2 -

Supplement to the Reload Safety Analyses, November 17, 1983.

8. Letter, Counsil to Miller, Millstone Nuclear Power Station, Unit No. 2 Cycle 7 Refueling - Preliminary Reload Safety Analysis - Proposed Revisions to Technical Specifications, February 6, 1985.

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TABLE 1 Millstone Unit 2 Cycle 7 Core Loading Initial BOC**

Number of Enrichment  % Theoretical Burnup Average Region M Assemblies w/o U235 Density (WWD/MTU)

A CE 4 1.93 95.0 15960 F1 y 4 2.70 94.5 25200 F2 g 5 3.30 94.9 22200 F2 g* 3 3.30 94.9 21560 G1 y 19 2.72 95.0 23470 G2 y 32 3.19 94.7 19290 G2 M* 4 3.19 94.7 9970 H1 g 30 2.73 95.2 13790 H2 M 44 3.22 94.8 9560 J1 W 24 2.62/2.91 95.2/95.1 0 J2 .H 48 2.91/3.29 95.1/95.2 0 i

Westinghouse fuel reassembled using CE skeletons.

• • E0L Cycle 6 burnup assumed: 11,500 MWD /MTU.

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n TABLE 2 MILLSTONE UNIT 2 KINETICS CHARACTERISTICS Current Limit Cycle 7 Most Positive Moderator. Temperature Coefficient (ap/'F) x 10 ~4 +0.5 from 0 to 70% Power +0.5 from 0 to 70% Power

+0.4 from 70 to 100% Power +0.4 from 70 to 100% Power Most Negative Moderator Temperature Coefficient (ap/*F) x 10~4, ARI -3.8 -3.8 Doppler Temperature Coefficient (Ap/F x 10-5)

-1.2 to -1.92 -1.2 to -1.92 Delayed Neutron Fraction B,ff .479 to .634 .479 to .640 Prompt Neutron Lifetime (usec) <32.2 <32.2 Maximum Differential Rod Worth cf two CEA groups moving together at HZP (pcm/in) 36.6 36.6 nenc-"*5 12

TABLE 3 SHUTDOWN REQUIREMENTS AND MARGINS MILLSTONE UNIT 2 - CYCLE 7 Control Rod Worth (%Ap) Cycle 6 Cycle 7 BOC EOC BOC EOC All Rods Inserted 7.49 8.37 7.83 8.60 All Rods Inserted Less Worst Stuck Rod 6.51 6.67 6.64 6.95 (1) Less 10 Percent 5.86 6.00 5.98 6.26 Control Rod Requirements (%Ap)

Reactivity Defects (Combined Doppler, T,yg,-Void and Redistribution' Effects) 1.86 2.64 1.67 2.58 Rod Insertion Allowance 0.52 0.38 0.87 0.41 (2) Total Requirements 2.38 3.02 2.54 2.99 Shutdown Margin ((1) - (2)) (%Ap) 3.48 2.98 3.44 3.27 Required Shutdown Margin (%Ap) 2.90 2.90 2.90 2.90 non em.'

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TABLE 4 SEQUENCE OF EVENTS - LOSS OF COOLANT FLOW Four pumps in operation, all pumps coasting down Event' Time (sec)

Loss of power to all pumps 0.0 Reactor coolant pump low .91 speed setpoint reached CEA's begin to drop 1.56 Minimum DNBR occurs 3.4 j.

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FIGURE 1 CORE LOADING PATTERN MILLSTONE UNIT 2 - CYCLE 7 A B C D E F GHJKLMNPR S T V W X Y  !

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U~ G2 F2 H2 F1 F2 G2 Gl 1 G2 G2 G2 F2 F1 H2 F2 G2 12 -

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F2 W 8 3.30 G1 W 19 2.72 G2 W 36 3.19  ;

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