ML20244B482
| ML20244B482 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/30/1989 |
| From: | Fetterman R, Foley J, Savage M WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20244B480 | List: |
| References | |
| NUDOCS 8904190222 | |
| Download: ML20244B482 (28) | |
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RELOAD SAFETY EVALUATION MILLSTONE UNIT 3 CYCLE 3 April 1989 Edited by: J. V. Foley Contributors: M. G. Savage P. W. Rosenthal R. J. Fetterman M. R. Wengerd A. J. Friedland ' L. G. Pilgrim C. A. Bly Approved: i E. H. Novendstern, Manager [: Thermal Hydraulic Design & Fuel Licensing C ammercial Nuclear Fuel Division Io ): Westinghouac Electnc Ccx1xration r'- % Nuclear Fuel Diviskm P. O. Ikm 3912 Pittsburgh Pennsyhrnia 15230 ATTACllMENT To TIIIL89-211
TABLE OF CONTENTS
1.0 INTRODUCTION
AND
SUMMARY
. . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1.1 - INTRO D UCTION . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.2 GENERAL DESCRIPTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 13 CO NCLUSION S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.0 REACTOR DESIG N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.1 MECHANICAL DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.2 NUCLEAR DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 23 THERMAL AND HYDRAULIC DESIGN . . . . . . . . . . . . . . . . . . . . . . . . . 7 3.0 POWER CAPABILITY AND ACCIDENT EVALUATION . . . . . . . . . . 8 3.1 POWER CAPABILITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.2 ' ACCIDENT EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3.2.1 KINETICS PARAMETERS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 3.2.2 CONTROL ROD WORTHS .. . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.23 CORE PEAKING FACTORS . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 33 ACCIDENTS REANALYZED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 33.1 RCCA EJECTION ACCIDENTS . . . . . . . . . . . . . . . . . . . . . . . . . 10
. 33.2 STEAM SYSTEM PIPING FAILURE . . . . . . . . . . . . . . . . . . . . . 11 333 REACTOR COOLANT PUMP SHAFT SEIZURE . . . . . . . . . . . . 11 4.0 TECHNICAL SPECIFICATION CHANGES ...................13 5.0 REFERENC E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 l
l l I , { 1 l 4 i l i i
)
> n { . LIST OF TABLES
' FUEL ' ASSEMBLY DESION PARAMETERS -
a MILLSTONE UNIT 3 - CYCLE 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
' KINETICS CHARACTERISTICS .
MILL. STONE UNIT 3 - CYCLE 3 N AND N-1 LOOP OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 END-OF-CYCLE SHUTDOWN REQUIREMENTS AND MARGINS MILLSTONE UNIT 3 - CYCLE 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 PARAMETERS FROM THE REANALYSIS OF THE RCCA EJECTION EVENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 TIME SEQUENCE OF EVENTS FOR REANALYZED NON.LOCA EVENTS . . . . . . 20 LIST OF FIGURES CORE LOADING PATTERN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 K(Z) - NORMALIZED Fo(Z) AS A FUNCTION OF CORE HEIOHT FOR FOUR LOOP OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 K(Z) - NORMALIZED Fo(Z) AS A FUNCTION OF CORE HEIOHT FOR THREE LOOP OPERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 7 hk
MILLSTONE UNIT 3 CYCLE 3 APRIL 1989
1.0 INTRODUCTION
AND
SUMMARY
1.1 INTRODUCTION
l This report presents an evaluation for Millstone Unit 3 Cycle 3, which demonstrates that the core ) l reload will not adversely affect the safety of the plant for N or N 1 loop operation. This evaluation was accomplished utilizing the methodology described in WCAP-9273-A, " Westinghouse Reload Safety Evaluation Methodology"0) for the incidents presented in the FSAR(2) that are within Westinghouse's scope. ; l i This evaluation includes consideration for Technical Specification revisions that allow relaxation i i of the end of life moderator temperature coefficients) wmtit the use of Ag-In-Cd rod cluster control assemblies (RCCAs) and modify the RCP ur- pedo 3) and overtemperature AT (OTDT)04) reactor protection system setpoints. However, implementation of the Technical Specification revisions for the RCP underspeed and OTDT setpoints is not required for Cycle 3 operation. Based upon the above referenced methodology, only those incidents analyzed and reported in the FSAR(2) that are within Westinghouse's scope which could potentially be affected by this fuel reload have been reviewed for the Cycle 3 design described herein. The results of analyses for incidents reanalyzed and the justification for the applicability of previous results is provided. 1.2 GENERAL DESCRIPTION The Millstone Unit 3 reactor core is comprised of 193 fuel assemblies arranged in the core loading pattern configuration shown in Figure 1. The Cycle 3 core loading configuration features a low leakage pattern. During Cycle 2/3 refueling,32 fresh region 5A assemblics,44 fresh region 5B I assemblies and 9 Region 2 assemblies from the spent fuel pool will replace 45 Region 2 fuel assemblies and 40 Region 3 fuel assemblies. A summary of the Cycle 3 fuel inventory is given in Table 1. i 1
MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 Nominal core design parameters utilized for Cycle 3 are as follows: Four Loop (N) Three Loop (N-1) LOCA* NON-LOCA" Core Power (MWt) 3411 2217 2560
. System Pressure (psia)- 2250 2250 2250 Core Inlet Temperature (oF) 557.0 551.2 550.6 Thermal Design Flow (gpm) 378,400 298,800 298,800 Average Linear Power Density (kw/ft) 5.434 3.532 4.078
1.3 CONCLUSION
S From the evaluation presented in this report, it is concluded that the Cycle 3 design does not cause the previously acceptable safety limits for any incident to be exceeded. This conclusion is based on the following:
- 1. Cycle 2 burnup of 15,800 +@ MWD /MTU.
- 2. Cycle 3 burnup is limited to the end-of-life full power capability "*
(nominally 16,500 MWD /MTU) plus a 500 MWD /MTU power coastdown.
- 3. There is adherence to plant operating limitations given in the plant Technical Specifications.
- Plus I.ocked Rotor Rods in DNB l' " Not including Locked Rotor Rods in DNB
"* Definition: Full-rated power and temperature (approximately 587oF Tavg), with control rods fully withdrawn and 0 to 10 ppm residual boron.
t 2 })
l 1 MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 l 2.0 REACTOR DESIGN i 1 2.1 MECHANICAL DESIGN l The mechanical design of the Region SA and 5B fuel assemblies is the same as the Region 4 fuel assemblies except that the Region 5 assemblies will incorporate several upgraded fuel design features. These features include: (1) Extended Burnup Capability, (2) Reconstitutable Top Nozzles ! l (RTNs), (3) Debris Filter Bottom Nozzles (DFBNs), (4) Integral Fuel Burnable Absorbers (IFBA), (5) Axial Blankets, and (6) Snag-Resistant Grids. These design improvements, described below, meet all fuel assembly and fuel rod design criteria. (1) Extended Burnun Canability The Region 5 fuel assembly design was modified for extended burnups by reducing the thickness of both the top nozzle and bottom nozzle end plates, decreasing the height of the top nozzle and bottom nozzle, and increasing the fuel rod length with a corresponding increase in the length of the fuel rod plenum. The fuel assembly overall height was adjusted to be consistent with fuel assembly growth predictions based upon accumulated Westinghouse in-core experience. This experience includes the results of high burnup demonstration programs conducted jointly by Westinghouse and utilities. These design changes allow for an additional distance between the nozzle plates, which is allocated for two purposes: (1) increased fuel rod growth associated with extended burnup and (2) increased fuel rod length to add plenum space for the increased fission gas release that occurs with increased burnup. As part of this design change, the grid elevations were relocated slightly to standardize the 17x17 fuel assembly design. Analyses have indicated the acceptability of the mechanical integrity of all fuel assembly components for extended burnup levels with the above changes. The methods and criteria established for Westinghouse fuel at extended burnup have been justifiedW and approved by the NRC. 3 l
MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 (2) Reconstitutable Ton Nozzle (RTN) The RTN differs from the current design in two ways: a groove is provided in each thimble thru-hole in the nozzle plate to facilitate attachment and removal; and the nozzle plate thickness was reduced to provide additional space for fuel rod growth. In conjunction with the RTN, a .long tapered fuel rod bottom end plug is used to facilitate removal and reinsertion of the fuel rods. Details of the RTN design features, the design basis, and the evaluation of the RTN are given in Section 2.3.2 of Reference 3 which has been approved by the NRC. (3) Debris Filter Bottom Nozzle (DFBN) This bottom nozzle is designed to inhibit debris from entering the active fuel region of the core and thereby improves fuel performance by minimizing debris related ft.el failures. The DFBN utilizes the same material, geometry, and welding requirements as its existing bottom nozzle counterpart. The DFBN is a low profile bottom nozzle design made of stainless steel, with reduced plate thickness and leg height thus providing additional space for fuel rod growth as part of the extended burnup feature. The DFBN is hydraulically equivalent to the existing bottom nozzle and meets all mechanical design functional requirements. (4) Intecral Fuel Burnable Absorber (IFBA) The IFBA coated fuel pellets are identical to the enriched uranium dioxide pellets except for the addition of a thin boride coating on the pellet cylindrical surface along the central portion of the fuel stack length. IFBAs provide power peaking and moderator temperature coefficient control. Details of the IFBA design are given in Section 2.5 of Reference 3 which has been approved by the NRC. l (5) Axial Blankets The arial blanket consists of natural uranium (approximately 0.74 w/o) dioxide pellets at each end of the fuel stack to reduce neutron leakage and to improve uranium utilization. The axial blanket pellet design is the same as the enriched and IFBA pellet designs except for an increase in length. 'Ihe length difference in the axial blanket pellets will 4
MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 I l~ help prevent accidental mixing with the enriched and IFBA pellets. Axial blankets are
- further discussed in Sections 2.4 and 3.3 of Reference 3 which has received NRC l
l approval. (6) Snar-Resistant Grids The snag-resistant grids contain outer grid straps that are modified to help prevent assembly hangup due to grid strap interference during fuel assembly removal This was accomplished by changing the grid strap corner geometry and adding guide tabs on the outer grid strap. Intermediate vanes to the top and tabs to the bottom of grids to reduce the potential of an assembly overlapping (and possibly locking) onto an adjacent fuel assembly. The corner chamfer is formed in the outside strap punching operation'to eliminate grinding and the resultant sharp edge. In addition, a weld is placed on the small overlap on the top and bottom of the corners to increase strength and round over the leading edge of the corner. In Cycle 3, some of the Hafnium RCCAs may be replaced with Ag-In-Cd Enhanced Performance a Rod Cluster Control Assemblies (EP-RCCAs). The absorber diameter of the EP-RCCA is reduced
. slightly at the lower extremity of the rodlets in order to accommodate absorber swelling and minimize cladding interaction. However, the EP-RCCA design for Millstone Unit 3 does not include the wear resistance feature that is typically standard with the EP-RCCA design.
Table 1 presents a comparison of pertinent design parameters of the various fuel regions for i Cycle 3. The Region 5A and 5B fuel has been designed utilizing the latest Westinghouse fuel I performance modelW, the Westinghouse clad flattening model(s), and the Westinghouse extended burnup methodology. The Westinghouse fuelis designed and operated so that clad flattening will not occur for its planned residence time in the reactor. The fuel rod internal pressure design basis (7) is satisfied for all regions. I Westinghouse's experience with Zircaloy clad fuel is described in WCAP-8183, " Operational Experience with Westinghouse Cores (8). This report is updated annually. 5 , 1
MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 1 2.2 NUCLEAR DESIGN q The nuclear design of the Cycle 3 core used Westinghouse codes approved by the NRC and the standard calculational methods described in the Westinghouse Reload Safety Evaluation Methodology (0. This methodology is not affected by changes to the maximum uranium enrichment , l used in the fuel. The changes in physics characteristics for Cycle 3 are typical of the normal j i variations seen from cycle to cycle. ! The Cycle 3 core loading is designed to meet a Fo x P ECCS limit of 5 2.32 x K(Z)* for four loop operation and 5 2.60 x K(Z)* for three loop operation. The flux difference (AI) bandwidth during normal operation conditions is +3, -12 % for four loop operation and +5,-5 % for three loop operation. Table 2 provides a summary of changes in the Cycle 3 kinetics characteristics compared with .he current limit based on previously submitted accident analyses. The Cycle 3 values fall witFa the current limits with the exception of the least negative Dopper temperature coefficient. Table 3 provides the control rod worths and requirements at the most limiting condition during the cycle for a core of 61 Hafnium rod cluster control assemblies (RCCAs). The required shutdown margin is based on previously sul mitted accident analyses (4 The available shutdown margin , exceeds the minimum required. For Cycle 3 some Hafnium RCCAs may be replaced by Ag-In-Cd EP-RCCAs. This change has been evaluated to allow the exchange of the Ag-In-Cd EP-RCCAs for any number of Hf RCCAs provided that any control or shutdown bank consists entirely of only one type of absorber material. This is possible since both RCCA designs have similar neutronic characteristics. The largest change in total rod worth during the cycle is less than 100 pcm". Core peaking factors change by less than IE As a result, the core performance characteristics of the Ag-In-Cd EP-RCCAs remain I essentially the same. The available shutdown margin will exceed the minimum required shutdown
*K(Z) - See Figures 2 and 3.
"pcm = 10-5 hp 6
MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 7 margin and all other Technical Specification limits related to nuclear design will be met for any combination of Hf and Ag In-Cd RCCAs in the configuration (s) described above. The loading pattern for Cycle 3 shown in Figure 1. It contains a total of 5104 IFBA rods located in the Region 5 fuel assemblies. The locations of the IFBA and secondary source rods are shown in Figure 1. 1 2.3 THERMAL AND HYDRAULIC DESIGN No significant variations in thermal margins will result from the Cycle 3 reload. Sufficient DNB margin exists for all events to meet the design criteria (2,9) for the Cycle 3 reload core. The DNB core limits and safety analysis used for Cycle 3 are based on conditions given in Sections 1.0 and 3.0. Fuel temperatures were calculated using the revised thermal safety mode 100) and include the effects of standardized pellets. i l l I i l 7 ! J l
).
MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 l 3.0 POWER CAPABILITY AND ACCIDENT EVALUATION I l 3.1 POWER CAPABILITY The plant power capability for three and four loop operation has been evaluated considering the l L consequences of those incidents examined in the FSAR(2). It is concluded that the core reload will not adversely affect the ability to safely operate at 100 percent of rated thermal power (RTP) for four loop operation and 65 percent of RTP for three loop operation during Cycle 3. For the evaluation performed to address overpower concerns, the fuel centerline temperature limit of 4700oF can be accommodated with margin in the Cycle 3 core using the methodology described in j Reference 1. The time dependent densification mode 101) was used for these fuel temperature evaluations. The LOCA limit at rated power can be met by maintaining Foat or below 2.32 for four loop and 2.60 for three loop operation according to their normalized Fo envelopes (shown in Figures 2 and 3). 3.2 ACCIDENT EVALUATION The effects of the reload, including the mechanical design changes described in Section 2.1, on the design basis and postulated incidents analyzed in the FSAR and within Westinghouse *s scope were examined. The Boron Dilution transients for Modes 3 through 6 are performed by the utility with Westinghouse support. The use of Ag-In-Cd EP-RCCAs in Cycle 3 has also been evaluated with respect to the postulated licensing bases accidents. In most cases, it was found that the effects were accommodated within the conservatism of the initial assumptions used in the applicable safety analyses. The resolution of exceptions required the reanalysis of three FSAR Chapter 15 transients, discussed below. During Cycle 2 operation, analyses were performed to support Technical Specification revisions that change the RCP underspeed setpoint03) and the OTDT reactor protection setpoint04). The current non-LOCA evaluation includes consideration of these setpoint changes which were determined to be acceptable for the Cycle 3 reload. However, implementation of these revisions is not required for Cycle 3 operation. 8
MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 IFBA fuel was evaluated to determine the potential impact on the FSAR Large break LOCA analysis. The IFBA fuel rods have been shown by analysis to be less limiting than the FSAR analysis of record due to an inherent power density reduction caused by the neutron poisoning and flux depression of the absorber. IFBA power density reduction input to the ECCS evaluation n.odel has been verified to be conservative relative to the Millstone Unit 3 Cycle 3 core. The limiting fuel type assumption for the FSAR Large Break LOCA continues to apply. The IFBA fuel rod is bounded by the FSAR analysis of record. Thus, operation of Cycle 3 with IFBA fuel meets the requirements of 10CFR50.46 and Appendix K to 10CFR50. A core reload can typically affect accident analysis input parameters in the following areas: core kinetic characteristics, control rod worths, and core peaking factors. Cycle 3 parameters in each of these three areas were examined as discussed below to ascertain whether revisions to the accident analyses assumptions were required. 3.2.1 KINETICS PARAMETERS A comparison of the range of values encompassing the Cycle 3 kinetics parameters with the current limits is given in Table 2. All the kinetics values fall within the bounds of the current safety analysis limits except for the least negative Doppler Temperature Coefficient (N and N-1 Imop) shown in Table 2 and the EOL HZP rod ejection parameters shown in Table 4. The difference between the Cycle 3 least negative Doppler Temperature Coefficient and the current limit affects only one non-LOCA event, the inadvertent operation of the emergency core cooling system (ECCS) at power. This non-limiting event was the only non-LOCA event that did not assume a Doppler Temperature Coefficient bounded by the Cycle 3 timits. The impact of the revised Cycle 3 limit for least negative Doppler Temperature Coefficient on this event was evaluated. Based on the nature of the event and its sensitivity on the subject parameter, the determination was made that the conclusions of the FSAR remain valid. That is, inadvertent ECCS actuation during power operation presents no hazard to the integrity of the RCS. 9 l l
MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 l The EOL HZP rod ejection event was reanalyzed using kinetics parameters that bound Cycle 3 operrtion. These parameters are summarized in Table 4. The results of this reanalysis are l discussed in Section 33.1. 1 3.2.2 CONTROL ROD WORTHS Changes in control rod worths may affect differential rod worths, shutdown margin, ejected rod
' worths, and trip reactivity. Table 2 shows that the maximum differential rod worth of two RCCA control banks moving together in their highest worth region for Cycle 3 meets the current limit.
Table 3 shows that the Cycle 3 shutdown margin requirements are satisfied. 3.2.3 CORE PEAKING FACTORS Peaking factors for the dropped RCCA incidents were evaluated based on the NRC approved ! dropped rod methodology 2). Results for N and N-1 loop operation show that the DNB design basis is met for all dropped rod events initiated from full power. Peaking factors following control rod ejection are within the bounds of the current limits except for EOL HZP core conditions. The peaking factors for the misaligned rod and the reanalyzed steamline break have been evaluated and are within the bounds of the previous safety analysis limits. 3.3 ACCIDENTS REANALYZED 3.3.1 RCCA EJECTION ACCIDENTS The RCCA ejection accident initiated from hot zero power (HZP) conditions at end-of-life (EOL) was reanalyzed for Cycle 3, assuming the parameters described in Table 4. The HZP EOL RCCA ejection case was the only case reanalyzed because the existing assumptions and results for the other cases continue to bound Cycle 3 operation. It should also be noted that the HZP RCCA ejection event is only analyzed for N-loop operation since this analysis bounds the N-1 loop cases. TaHe 5 gives the time sequence of events for this accident. The results of the reanalysis, as indicated in Table 4, demonstrated that the conclusions of the FSAR for the RCCA ejection event remain valid. l' 10 l i
MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 3.3.2 STEAM SYSTEM PIPING FAILURE The main steam line rupture event for N-loop operation was reanalyzed for Cycle 3 using revised core kinetics parameters. Both the limiting case that assumes the availability of offsite power throughout the event and the less severe case that includes a loss of offsite power were reanalyzed. No reanalysis was required for the N-1 loop main steamline rupture, since the existing analysis for that case continues to bound Cycle 3. Table 5 provides the time sequence of events for both of the analyzed cases. A DNB analysis was performed for the limiting case and it was determined that the conclusions of the FSAR for the main steam line rupture event remain valid. That is the DNB design basis continues to be met for this event. 3.3.3 REACTOR COOLANT PUMP SHAFT SEIZURE (Locked Rotor) Reanalyses of the locked rotor event were performed for Cycle 3 to predict the numba of fuel rods that undergo DNB for N loop and N-1 loop operation. The analysis for N-1 loop operation assumed an initial nominal power level of 65% Rated Thermal Power (RTP). Previous analysis for this event, along with all other non.LOCA events, had assumed an initial nominal N-1 loop power level of 75% RTP. The actual licensed N-1 loop nominal power level is 65% RTP so that the use of 75% RTP represented a conservatism in the previous analysis. For Cycle 3, continued use of 75% RTP would have resulted in the predicted number of fuel rods that undergo DNB for this event exceeding the current limit value. The results of the locked rotor reanalysis verified that less than 8.0% of the fuel rods were
. predicted to undergo DNB for the N-1 loop locked rotor event with the 65% RTP assumption.
The reanalysis of the N loop locked rotor event verified that less than 6% of the fuel rods were predicted to undergo DNB. The radiological dose release evaluation for Cycle 3 is performed by the utility. The Cycle 3 locked rotor reanalysis described above was limited to the issue of determining the number of rods in DNB. The limiting cases of the current locked rotor licensing basis analysis t intended to predict other transient conditions such as maximum RCS pressure, maximum clad l l 11 I
l MIILSTONE UNIT 3 CYCLE 3 APRIL 1989 temperature, and the magnitude of the zirconium steam reaction remain valid for Cycle 3. It should be noted that for N-1 Loop, use of the current locked rotor analysis means continuing to use an initial N-1 loop nominal power of 75% RTP. The reduced N-1 loop nominal power level of 65% RTP was only used for the rods in DNB analysis. 12
i MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 ) 1 4.0 TECHNICAL SPECIFICATION CHANGES p. [- A Technical Specification amendment has been prepared which, if approved, will modify the RCP underspeedG3) and OTDWd) reactor protection system setpaints. This evaluation includes consideration for these revisions. Implementation of these Technical Specification changes will not affect the conclusions of this safety evaluation. ! l The Technical Specification changes required for Millstone Unit 3 Cycle 3 operation have been submitted to the NRC06). i l, l 13 l l
)
MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 i
5.0 REFERENCES
- 1. Davidson, S. L, et. al., " Westinghouse Reload Safety Evaluation Methodology,"WCAP-9273-A, July 1985.
- 2. " Final Safety Analysis Report Millstone Generating Station, Unit 3," USNRC Docket No. 50-423, December 1988.
i Davidson, S. L, et.al. " Reference Core Report VANTAGE 5 Fuel Assembly," 3. WCAP-10444-P A, September,1905.
- 4. Weiner, R. A., et. al., " Improved Fuel Performance Models for Westinghouse Fuel Rod Design and Safety Evaluations," WCAP-10851-P-A, August 1988.
- 5. George, R. A., et. al., " Revised Clad Flattening Model," WCAP-8381, July 1974.
- 6. Davidson, S. L, and Kramer, W. R., " Extended Burnup Evaluation of Westinghouse Fuel,"
WCAP-10125-P-A, December,1985.
- 7. Risher, D. H., et. al., " Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis," WCAP-8964-A, August 1978.
- 8. Foley, J., and Skaritka, J.; " Operational Experience with Westinghouse Cores,"
(through December 31,1987), WCAP-8183, Revision 16, August 1988.
- 9. Letter from A. C. Thadani (NRC) to W. J. Johnson (Westinghouse), Jan. 31,1989,
Subject:
Acceptance for Referencing of Licensing Topical Report, WCAP-9226-P/9227-NP, " Reactor Core Response to Excessive Secondary Steam Releases". l 10. Leech, W. J., et. al., " Revised PAD Code Thermal Safety Model," WCAP-8720, Addenda 2, October 1982. 14 l f
MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 l l 11. Hellman, J. M. (ed.), " Fuel Densification Experimental Results and Models for Reactor Operation," WCAP-8219-A, March 1975. l
- 12. Morita, T., et. at, " Dropped Rod Methodology for Negative Flux Rate Trip Plants,"
WCAP-10298-A, March 1983.
- 13. Westinghouse letter from D. L Fuller to J. A. Camp dated December 28,1988, NEU 719, " Northeast Utilities Service Company, Millstone Nuclear Power Station Unit No. 3.
RCP Underspeed Setpoint Analysis," (Proprietag).
- 14. Westinghouse letter from D. L Fuller to J. A. Camp dated December 30,1988, NEU 720, " Northeast Utilities Service Company, Millstone Nuclear Power Station Unit No. 3, Overtemperature Delta-T Setpoint Analysis," (Proprietary).
- 15. Fetterman, R. J., et. al., " Safety Evaluation Supporting a More Negative EOL Moderator Temperature Coefficient Technical Specification for the Millstone Nuclear Power Station Unit 3," WCAP-11946, September 1988.
- 16. Letter from E. J. Mroczka (Northeast Utilities Service Company) to NRC " Proposed Revision to Technical Specifications", January 24,1989.
i l 15
MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 TABLE 1 FUEL ASSEMBLY DESIGN PARAMETERS
. MILLSTONE UNIT 3 - CYCLE 3 Region 2 1 4A 4B 1A EH Enrichment (w/o U-235)* 2.899 3.395 3.497 3.808 4.10 " 4.50" Geometric Density
- 94.% 5 94.980 95.13 95.17 95 95
(% theoretical) Number of Assemblies 9 24 56 28 32 44 Approximate Burnup at 21,350 24,160 19,470 15,450 0 0 Beginning of Cycle 3 (MWD /MTU)*" i l [ . All values are as-built except Region SA and 5B
" Enrichment of enriched axial region of assemblies. Each Region 5 assembly also i six inches of 0.74 w/o axial blanket fuel at top and bottom of the assembly. {
I
'" Based on actual EOC1 burnup of 18,700 MWD /MTU and nominal EOC2 burnup of 15,800 MWD /MTU.
16
.:: MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 TABLE 2 KINETICS CHARACTERISTICS
[. MILLSTONE UNIT 3 - CYCLE 3 l N AND N-1 LOOP OPERATION 1 l~ Cycle 3 Changes Current Limit to Current Limits Most Positive Moderator Temperature +5,s 70% RTP -- Coefficient (pcmrF)* +5 to O Ramp from 70% to 100% RTP Most Negative Unrodded Moderator -47.5 -- Temperature Coefficient (pcmrF)* Doppler Temperature Coefficient, -2.9 to -1.4 -2.9 to -1.0 (pcmrF)* Least Negative Doppler - Only Power -10.18 to -6.68 - Coefficient, Zero to Full Power (pcm/% power)* Least Negative Doppler - Only Power -970 (4 Loop) - Defect at BOC (pcm)* -650 (3 Loop) -- Most Negative Doppler - Only Power -19.4 to -12.6 -- Coefficient, Zero to Full Power (pcm/% power)* Delayed Neutron Fraction Bert,(%) 0.44 to 0.70 -- (4 Loop) (4 Loop) 0.44 to 0.75 -- (3 Loop) (3 Loop) Maximum Differential Rod Worth to 77.33 -- Two Banks Moving Together at HZP (pcm/in)* pcm = 10 5 Ap 17 j
l-MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 TABLE 3 5 END-OF-CYCLE SHUTDOWN REQUIREMENTS AND MARGINS MILLSTONE UNIT 3 - CYCLE 3 4 Loop (N) 3 Loop (N 1) Cycle 3 Cvele 3 Control Rod Worth (%Ao) All Rods Inserted 7.61 7.61 All Rods Inserted Less Worst 6.48 6.48 Stuck Rod (1) I.ess 10% 5.83 5.83 Control Rod Requirements (%Ao) Reactivity Defects (Combined Doppler, 3.57 2.85 Moderator Temperature, Void) Rod Insertion Allowance 0.33 0.50 (2) Total Requirements 3.90 3.35 Shutdown Margin (1)-(2) (%Ap) 1.93 2.48 Requirement Shutdown Margin (%Ap) 1.60 1.60 f 4 l l-L i l l I 18
. MILISf0NE UNIT 3 CYC1,E 3 APRIL 1989 TABLE 4 PARAMETERS FROM THE REANALYSIS OF THE RCCA FJECTION EVENT j
HOT ZERO POWER AT END-OF-LIFE CASE ONLY ? Cvele 2 Cycle 3 Power Level (%) 0.0 0.0 Ejected Rod Worth (% Ak) 0.90 0.88 Delayed Neutron Fraction (%) 0.44 0.44 Feedback- Reactivity Weighting 3.55 3.625 Trip Reactivity (% Ak) 2.0 2.0 Fo After Rod Ejection 20.0 26.0 Number of Operational Pumps 2 2 Max.- Fuel Pellet Average Temperature (oF) 3290 3524 Max. Fuel Pellet Center Temperature (oF) 3675 3934 Max. Clad Average Temperature (oF) 2479 2678 Max. Fuel Stored Energy (cal /gm) 138 150 19 l l l J
' MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 1
1 TABLE 5 TIME SEQUENCE OF EVENTS FOR REANALYZED NON-LOCA EVEN'IS Time Accident Event (sec) N-Loop RCCA- Initiation of rod ejection 0.0 Ejection, HZP & EOL Power range high neutron 0.18 flux low setpoint reached Peak nuclear power occurs 0.21 Rods begin to fall into core 0.68 Peak clad temperature occurs 1.32 Peak heat flux occurs 1.36 Peak average fuel temperature 1.52 N-Loop Steam System Piping Failure
- 1. With off-site Steam line ruptures 0.0 power available Pressurizer empty 14.6 Criticality attained 30.4 Injected boron reaches core 72.8
- 2. With loss of Steam line ruptures 0.0 off-site power Pressurizer empty 16.6 Criticality attained 38.4 Injected boron reached core 86.8 l
i 20 l
' MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 L
L FIGURE 1 MILLSTONE UNIT 3 CYCLE 3 CORE LOADING PATTERN 5104 IFBA IN 84 ASSEM8 LIES R P' - N M L K. J H G F E D C 8 A 180 l 4A I 58 48 58 48 58 4A
'l
! 0' 0 0
~
4A 58 4A 58 4A 58' 4A 58 4A l /,8 l 48
!. 32 100 100 32 I 48 i SA 58 4A 5A 3 g 3 5A 4A 58
.00 5A 32 48 1 32 100 80 80' 4A ; 58 -3 5A' 2 58 I 4A 58 2 5A 3 58 4A i 80 100 80 100l 100 100 4A 5A 4A 48- 4A .3 4A 48 4A 5A I 4A 58 4A I 58 l 4A 32 80 i 80 1 32
. 58 4A i , 5A - 2 48 4A 5A 3 5A 4A 48 2 1 5A 4A 58 0 i'80 i 80 80 l 80 0 5A 5A 4A dB 7
48 58.] 3 58 4A 5A 3 3 58 l3 5B
-100 1 100 80 100 80 100 l 100 o SB 4A i 48 4A 3 3 SA l 2 5A 3 3 4A 48 4A 58 o 90 I 0 100 1 100 0 48 58 4A BA 3 5A 3 5A 4A 5B l3 58 48-
, SB l 3 901 100 80 100 80 100 l 100 58 - 4A I SA 2 48 4A I SA 3 SA 4A 48 2 5A I 4A BB 0 ! 80 - l 80 80 80 l 0 5B + 4A SA- i 3 4A 48 4A SA l4A 58 4A l 4A' 4A 4 48 l 4A l 32 80 l 1 l 80 l 32 4A i 5B 3 5A 2 58 4A 5B 2 SA 3 58 4A I 100 80 100 100 80 100 48 . ! 5A 58 4A 5A 3 g 3 5A 4A 5B 5A 32 48
' 32 100 80 , 80 100 48 4A SB 4A SB t 4A 58 4A 58 4A 48 32 100l 100 32 i 4A 58 48 58 4A 58 l 48 0l 0 0 0'
SFik N dDS 21
MILLSTONE UNIT 3 CYCLE 3 APRIL 1989 FIGURE 2 K(Z) - NORMALIZED Fo(Z) AS A FUNCTION OF CORE HEIGHT FOR FOUR LOOP OPERATION 1.2 I I l(0, 1.00)l I(6. 1.00)l 1.0 10.8, .94) m
.80
_ l(12, .647)l
\
C1.60 x
.40
.20 l
t l' O.00 0 2 4 6 8 10 12 l-l CORE HEIGHT (feet) 22 4
MILLSTONE UNrr 3 CYCLE 3 APRIL 1989 FIGURE 3 K(Z) - NORMALIZED Fo(Z) AS A FUNCTION OF CORE HEIGHT l FOR THREE LOOP OPERATION 1.2 l (0, 1.00) l l (6. 1.00) 1.0
^%
l ( 10. 8, .94) l
.80
\
l(12,.577)L S.60 '
)
x
.40
.20 0.00 12 CORE HEIGHT (feet {
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