ML19323H680

Reload Safety Analysis.

ML19323H680
Person / Time
Site:	Millstone
Issue date:	06/30/1980
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML19323H679	List: ML19323H678 ML19323H680
References
TAC-11348, TAC-11561, TAC-12505, TAC-42846, NUDOCS 8006160048
Download: ML19323H680 (23)

Text

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DOCKET NO. 50-336 O

ATTACliMENT MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 RELOAD SAFETY ANALYSIS i

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JUNE, 1980

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TABLE OF CONTENTS Section Title Pjgte

1.0 INTRODUCTION

AND

SUMMARY

1 1.1 OBJECTIVES 1 1.2 GENERAL DESCRIPTION 1

1.3 CONCLUSION

S 2 2.0 MECHANICAL DESIGN 3 2.1 GENERAL DISCUSSION 3 2.2 GUIDE TUBE INSET DEMONSTRATION PROGRAM 3 3.0 THERMAL AND HYDRAULIC DESIGN 5 4.0 NUCLEAR DESIGN 6 5.0 ACCIDENT ANALYSIS

5.1 INTRODUCTION

AND

SUMMARY

7 5.2 ACCIDENT EVALUATION 7 5.2.1 KINETICS PARAMETERS 8 5.2.2. CEA WORTHS 8 5.2.3 CORE PEAKING FACTORS 8 5.3 ' INCIDENTS REANALYZED 8 5.3.1 BORON DILUTION 8 5.3.2 CEA EJECTION INCIDENT 9 Appendix A References A-1 l

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LIST OF TABLES Table Title Page 1

Millstone Unit 2 Cycle 4 Core Loading 10 2 Millstone Kinetics Characteristics 11 3 Shutdown Requirements and Margins 12 4 Parameters Used in the Analysis of the 13 CEA Ejection Accident 5 Sequence of Events, CEA Ejection Incident - HZP 14 4

i LIST OF FIGURES Figure Title

, Page 1 Cycle 4 Loading Pattern 15 2 Westinghouse Guide Tub'e Insets '

16 3 Millstone 2 - Safety Analysis Rod Ejection 17 Incident - HZP/BOL Nuclear Power Versus Time 4 Millstone 2 - Safety Analysis Rod Ejecticn 18 Incident - HZP/BOL Fuel and Clad Tempera'ture Versus Time b

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1.0 INTRODUCTION

AND SU N RY 1.1 OBJECTIVES This report presents an evaluation for Millstone Nuclear Power Station Unit 2, Cycle 4, 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 4 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 The Millstone 11 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) Inconel grids, a stainless steel sottom 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 dio~xide 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 4 are as follows:

CorePower(Mwt) 2700 System Pressure (psia) 2250 Reactor Coolant Flow (GPM) 370,000*

Core Inlet Temperature (OF ) 549 l

Average Linear Power Density (kw/ft) 6.065 (based on best estimate hot, densified core average stack height of 134.4 inches)

• Minimum guaranteed safety analysis flow ,

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The core loading pattern for Cycle 4 is shown in Figure 1. Twenty-four (24) interior feed assemblies containing 2.7 w/o U235 and f'orty-eight (48) peripheral feed assemblies containing 3.3 w/o U?35 are replacing sixty-eight (68) Combustion Engineering (CE) batch C assemblies and four (4) CE batch B assemblies. The fifth batch B assembly in Cycle 3 is replaced by a batch B assembly that was discharged at the end of Cycle 1.

A summary of the Cycle 4 fuel inventory is given in Table 1.

1.3 CONCLUSION

S From the evaluation presented in this report, it is concluded that the Cycle 4 design does not result in the previously acceptable safety limits for any incident to be exceeded. This conclusion is based on the following:

1. Cycle 3 expected burnup of 10,250+{3@ MWD /MTU.

2. There is adherance to plant operating limitations'as given in the Technical Specifications.

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2.0 MECHANICAL DESIGN 2.1 GENERAL DISCUSSION A full description of the mechanical design of the Westinghouse Millstone II reload fuel assembly to be utilized in Cycle 4 is given in the BSR. As discussed in the BSR, the Westinghouse fuel assemblies have been designed to be compatible with the resident Millstone II fuel assemblies.

2.2 GUIDE TUBE INSET DEMONSTRATION PROGRAM A four assembly demonstration program utilizing a guide tube inset design is planned for Cycle 4. The Westinghouse inset design (see Figure 2) is a structural modification made to the fuel assembly guide tubes that is intended to greatly reduce the guide ' ube wearing condi-tion by standing off,the tip of the control rod fesm the guide tube wall, thereby reducing the amplitude of rod vibration. These insets are 1.32 inch long by .33 inch wide rectangular deformations that reduce the original tube diameter locally from 1.035 inches to .980 inches. Four individual insets are located at two axial elevations of the guide tubes.

Results of initial tests performed to date demonstrate that the inset design provides advantages over the existing sleeve design in reducing wear. These advantages are as follows:

1. The locally reduced diametral clearance limits the lateral motion of the tip and also reduces the impact loading of the tube due to the rodlet motion.

2. The insets are located above the region of the rodlet where the point contact occurs which forces the rod into more of a line contact wearing mode of the inset.

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3. The four fnset geometry influences the rod to a two line support which is a more stable state of equilibrium than the' single point sphere on cylinder or single line cylinder on cylinder contact.

This is a much more favorable condition for wear for the support of .

a stiff rod.

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3.0 THERMAL AND HYDRAULIC DESIGN A description of the thermal and hydraulic design of the Westinghouse Millstone II reload fuel assembly to be utilized in Cycle 4 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 the resident Millstone II fuel assemblies.

Steady state DNB analyses of Cycle 4 have been performed using the THINC-I code with the W-3 DNB correlation. In these analyses, the assumed uncertainties in various measured parameters are factored in as biases. This biasing of the measurement uncertainties in the analyses is equivalent to adding the absolute power uncertainties equivalent to these uncertainties in the various measured parameters and applying the total power uncertainty to the best estimate calculation. The specific' uncertainties along with their equivalent power uncertainties as determined with THINC-I and the W-3 DNB correlation are given below:

Equiv. Power

! Parameter Uncertainty Uncertainty ASI 0.06 ASIU '3.0%

Pressure 22 psi 0.5%

T in 2F 1.0%

Flow 4% 2.0%

Power (LCO) 2% 2.0%

Power (LSSS) 5% 5.0%

In the Cycle 4 analysis, the equivalent sum of these uncertainties is 8.5% for LCO and 11.5% for LSSS. Treating these uncertainties as sta-tistically independent, the Root Sum 5quare method is used in combining them. This RSS combination yields 4.3% for LC0 and 6.3% for LSSS, giving a net conservatism of 4.2% for LCO and 5.2% for LSSS. A partial credit of 3.0% is taken for both the LC0 and LSSS., .

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4.0' NUCLEAR DESIGN The Westinghouse nuclear design procedures, computer programs, and calculation models utilized in the Millstone II, Cycle 4 reload design are presented in the BSR. With one exception, no changes or modifica-tions were required for Cycle 4 evaluation. The exception relates to the local power density trip setpoint methodology. Cycle 4 accident simulations take credit for the variable high power trip by tenninating 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.

i The Cycle 4 core loading results in a maximum linear heat rate of less than 15.6 kw/ft at all fuel heights at rated power. Table 2 provides a sunnary of changes in the Cycle 4 kinetics characteristics compared with the current limit based on the reference safety analysis (2) . It can be seen from the table that all of the Cycle 4 values fall within current limits, except as discussed in Section 5.3. Table 3 provides the contol rod worths and requirements at the most limiting condition' during the cycle. The required shutdown margin is based on previously submitted accident analyses (2) . The available shutdown margin exceeds the minimum required.

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5.0 ACCIDENT ANALYSIS

5.1 INTRODUCTION

AND

SUMMARY

i 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. It is concluded that the core reload will not adversely affect the ability to safely operate at 100% of r,ated power during Cycle 4. For the overpower transient, the fuel centerline temperature limit of 47000F can be accomodated with margin in the Cycle 4 core. The burnup dependent densification model(3,4) was used for fuel temperatum evaluations. The LOCA limit at rated power can be met by maintaining Fg at or below 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 DSR(2) were examined. In most cases, it was found that the effects were accommodated within the conservatism of the initial assumptions used in the BSR safety analysis. For those incidents which were reanalyzed, it was determined that the applicable ,

design bases are not exceeded, and, therefore, the conclusions presented in the BSR are still valid.

A core reload can typically affect accident analysis input parameters in the following areas

core kinetic characteristics, CEA worths, and core peaking factors. Cycle 4 parameters in each of these areas were examined as discussed below to ascertain whether new accident analyses were required.

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5.2.1 KINETICS PARAMETERS A comparison of Cycle 4 kinetics parameters with the current limits, established by the BSR safety analyses, is presented in Table 2. All parameters in Table 2 except the Doppler coefficient, were found to be within the limiting range of values used in the BSR safety analyses.

5.2.2 CEA WORTHS Changes in CEA worths may affect shutdown margin. T.able 3 shows that the Cycle 4 shutdown margin requirements are satisfied.

5.2.3 CORE PEAKING FACTORS Peaking factor evaluations were performed for rod out of position, and steam line break accidents to ensure that the DNB design limits are not exceeded. These evaluations were performed'util.izing the existing transient statepoint information from the reference cycle (BSR) and peaking factors determined for the reload core design. In each case, it was found that the peaking factor for Cycle 4 yielded results that were within the DNB design limits. Consequently, for these accidents no further investigation or analysis was required.

CEA re' king factors for Cycle 4 were within the reference cycle limits.

5.3 INCIDENTS REANALYZED 5.3.1 BORON DILUTION Critical baron concentrations for Cycle 4 were conservatively deter-mined to be higher than the critical boron concentrations of Cycle 3.

Therefore, the boron dilution incident was reanalyzed for those cases where the Cycle 4 critical boron concentrations were higher.

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Critical Baron Concentration (ppm)

Cycle 3 Cycle 4 Power Operation 1300 1363 Startup 1400 1453 Hot Standby 1400 1453 Hot Shutdown 1400 1407 Refueling 1300 1383 For all cases, there was sufficient time for operator action to terminate the baron dilution before shutdown margin is lost. In the power operation case, the reactivity addition rate is much smaller than that due to rod withdrawal.

Results Power Operation 4.35 x 10-6 ap/sec Startup 53 min. to lose shutdown Hot Standby 42 min.

Hot Shutdown 42 min.

Refueling 34 min.

5.3.2 CEA EJECTION INCIDENT The hot zero power case was re-analyzed for Cycle 4 because the Fq af ter CEA ejection is higher than the Fq used in the BSR analysis. The calculated fuel temperature and the maximum fuel stored energy are higher; but no fuel melting occurs.

Table 4 gives the key analysis parameters assumed in the analysis.

The sequence of events for this accident is given in Table 5. The nuclear power transient and hot spot fuel and clad temperature tran-

sients are shown in Figures 3 and 4.

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TABLE 1 Millstone Unit 2 Cycle 4 Core Loading Initial BOC ,

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

B+ CE 1 2.336 95 17566 D1 CE 24 2.7349 94.75 21363 02 CE 48 3.0207 94.75 19380

, El CE 24 2.730 94.75 12759 l E2 CE 48 3.235 94.75 8829 F1 E. 24 2.70 95* 0 i F2 g 48 3.30 95* 0

• The Region F1 and F2 densities are nominal. Average dens'ities of 94.5% theoretical were used for Region F1 and F2 nuclear design

evaluations.

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. TABLE 2 .

MILLSTONE KINETICS CHARACTERISTICS Current Limit (BSR) Cycle 4 Moderator Temperature Coefficient (Ap/ 0F) x 10-4 +0.5 to -3.8 +0.5 to -3.8 Doppler Coefficient (Ap/0F x 10-5) -1,2 to -1.9 -1.2 to -1. 9 ,

Delayed Neutron Fraction Beff(%) .479 to .624 .479 to .624 Prompt Neutron Lifetime (psec) 32.2 <32.2 Maximum Differential Rod Worth of two CEA groups moving together at HZP (pcm/in) 24.3 <24.3 e

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TABLE 3 SHUTDOWN REQUIREMENTS AND MARGINS MILLSTONE UNIT 2 - CYCLE 4 Control Rod Worth (%Ao) Cycle 3

• Cycle 4

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BOC EOC BOC E0C All Rods Inserted 9.70 11.00 8.24 9.11 All Rods Inserted Less Worst Stuck Rod 6.60 7.50 6.78 7.02 (1) Less 10% 5.94 6.75 6.10 6.32 Control Rod Requirements Reactivity Defects (Combined Doppler, Tayg, Void and Redistribution Effects) 1.80 2.30 1.71 2.62 Rod Insertion Allowance 0.60 0.60 0.36 0.36 (2) Total Requirements 2.40 2.90 2.07 2.98 Shutdown Margin ((1) - (2)) (%Ap) 3.54 3.85 4.03 3.34 Required Shutdown Margin (%Ap) 3.20 3.20 3.20 3.20

• Cycle 3 values from Reference 5 ,

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TABLE 4 PARAMETERS USED IN THE ANALYSIS OF THE CEA EJECTION ACCIDENT Parameter HZP Power Level, 0 i

Ejected rod worth, AK 0.65 Delayed neutron fraction, 0.47 Feedback reactivity weighting 2.50 Trip reactivity, Ak 2.1 Fq before rod ejection --

F after rod ejection q 18.8 Number of operational pumps 2 Results Max. fuel pellet average temperature, F 3346 Max. fuel center temperature, OF 3845 Max. clad average temperature, OF 2528 Max. fuel pellet center melting, percent 0 Max. fuel stored energy, cal /gm 141 0

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TABLE 5 SEQUENCE OF EVENTS, CEA EJECTION INCIDENT - HZP Time Event Setpoint or Value 0.0 Initiation of Transient -

0.1 -

CEA Fully Ejected -

0.28 High Power Trip Signal Generated 25 percent 0.35 Peak Nuclear Flux Reached See Fig. 3 1.18 CEA Insertion Begins -

2.84 Peak Fuel Temperature Reached See Fig. 4 3.78 CEA's Reach 90 percent Insertion -

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G .H JK L MN P R A B C D E F I l  ! I S T V W X Y F2 F2 F2 F2 1

, F2 F2 F2 El E2 El F2 F2 F2 F2 D2 D2 D2 3

F1 D2 D2 E2 D2 F2 3

F2 F1 E2 4

D1 F1 D2 D1 D2 F1 D1 E2 F1 F2 F2 E2 El 5 l D2 E2 D1 E2 E2 El E2 D1 E2 E2 D2 F2 F2 E2 D1 D1 R D2 E2 El E2 D2 F1 D1 D1 E2 F2 6 F2 D2 F1 E2 D2 E2 D2 F1 D2 E2 D2 E2 F1 02 F2 F2

_g El D2 D2 El E2 D2 El D1 El D2 E2 El D2 D2 El F2 F2 - 10 F1 D1 E2 El F1 D1 D1 F1 El E2 D1 F1 E2 -- 1 1 F2 3 F2 - 12 El D2 D2 El E2 D2 El D1 El D2 E2 El' 02 D2 El 13

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, F2 D2 F1 E2 D2 E2 D2 D2 E2 15 F1 D2 E2 F1 02 F2 F2 E2 D1 D1 F1 D2 E2 El E2 D2 F1 D1 D1 E2 F2 16 F2 D2 E2 E2 D1 E2 El E2 El E2 Di E2 E2 D2 F2 F2 F1 E2 D1 F1 D2 D1 D2 F1 D1 E2 F1 F2 18 i E2 F2 D2 E2 D2 D2 F1 D2 D2 02 F2 19 ss F2 F2 F2 El E2 El F2 F2 F2 20 F2 F2 F2 21 F2l kgi n h w/o U235 REACTOR CORE B+ CE 2 .336 D1 CE 2.735 SS - Secondary Source D2 CE 3.021 BP - Burnable Poison El CE 2.730 E2 CE 3.2 35 F1 W 2.70 F2 W 3.30 Fi g.1 Cycle 4 Loading Pattem

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Figure 3 Mil 1 stone 2 - Safety Ana1ysis ~ Rod Ejection Incident - HZP/BOL Nuclear Power Versus Time f

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Figum 4 Millstone 2 - Safety Analysis Rod Ejection Incident - HZP/BOL Fuel.

and Clad Temperatures Versus Time

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APPENDIX A I

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REFERENCES

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

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 Experiemental Results and Model for Reactor Operation", WCAP-8219-A, March 1975.

5. Letter, Counsil to Reed, Millstone Nuclear Power Station, Unit No. 2, Proposed License Amendment, Power Uprating, February 12,1979 e

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