ML20211Q317
| ML20211Q317 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 09/09/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20211Q315 | List: |
| References | |
| NUDOCS 9909150042 | |
| Download: ML20211Q317 (12) | |
Text
,
!paug4 UNITED STATES 1
[o
,j NUCLEAR REGULATORY COMMIS 'ON WASHINGTON, D.C. 2066H001
- s. ... /
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGQLATION l ACCEPTANCE OF BOILING WATER REACTOR OWNERS' GROUP REPORT.
" PREDICTION OF THE ONSET OF FISSION GAS RELEASE I
FROM FUEL IN GENERIC BWR."
l ENTERGY OPERATIONS. INCORPORATED GRAND GULF NUCLEAR STATION (GGNS)
DOCKET NO. 50-416
1.0 INTRODUCTION
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By letter dated May 6,1997, Entergy Operations, Inc. (the licensee), requested NRC review for acceptance of a report," Prediction of the Onset of Fission Gas Release From Fuelin Generic BWR," dated July 1996 (Reference 1). This report was prepared by the General Electric Company (GE) and was sponsored by the Boiling Water Reactor Owners' Group (BWROG).
This report (BWROG report) documents the results of an analysis performed to determine the ,
minimum time to the onset of fission product release from perforated fuel rods following a I postulated design-basis loss-of-coolant accident (LOCA). GE calculated the minimum time to the onset of fission product release from perforated fuel rods to be 121 seconds using a bounding BWR plant configuration and fuel design. This calculation is intended to be generic for all currently operating BWR p! ants using currently licensed BWR fuel.
In NUREG-1465," Accident Source Terms for Light-Water Nuclear Power Plants" (Reference 2), the staff estimated the minimum time to the onset of fission product release from I perforated fuel roos to be about 30 seconds unless plant-specific calculations are performed. !
The licensee specifically requested that the information in the BWROG report be reviewed to replace the minimum time for fission product release currently given in NUREG-1465.
2.0 BACKGROUND
in SECY-96-242,"Use of the NUREG-1465 Source Term at Operating Reactors," dated November 25,1996, the staff informed the Commiuhn of its approach to allow the use of the revised accident source term described in NUREG-1465 at operating plants. In the SECY paper, the staff also described its plans to (1) undertake a rebaselining assessment of two nuclear power plants to further evaluate the issues involved with applying the revised accident source term at operating plants, (2) review the pilot plant applications implementing the revised 4 accident source terms following completion of the rebaselining effort, and (3) incorporate the total effective dose equivalent methodology in the review of the pilot plant applications. The Commission approved these plans and directed the staff to commence rulemaking upon completion of the rebaselining and concurrent with the pilot plant reviews.
9909150042 990909 PDR ADOCK 05000413 P PDR
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2 The aff has completed its rebaselining effort and presented the results in SECY-98-154. The staff SUE Htted a rulemak!ng plan for implementation of the revised source term at operating reactors in t 9CY 98-158. In response to this plan, the Commission directed the staff to allow limiteu or seiective application of the revised source term at operating reactors and promptly complete review of the pilot plant initiatives. Accordingly, the staff has initiated review of submittals from pilot plants.
GGNS is a lead pilot plant requesting selective implementation (fission product release timing only) of the revised ac;ident source term presented in NUREG-1465. In addition to GGNS, the
! following pilot plants submitted license amendments using the revised accident source term:
Perry Unit 1, Browns Ferry Units 1,2, and 3, Indian Point Unit 2, and Oyster Creek.
3.0 EVALUATION GE performed its calculations using LOCA methodology that had been previously reviewed and approved by the staff. In this methodology, the SAFER computer code calculates the long-term system response of the reactor. The CHASTE computer code is then used to model fuel rod heatup for the highest power axial plane in the highest power fuel rod assembly. GE used this approved methodology to evaluate the minimum time tc iW oerforation and to perforrn sensitivity studies in determining the most limiting BWR vesse: design, fuel rod design, and core burnup.
As further verification of the minimum time to the onset of fission product release from perforated fuel rods following a postulated design-basis LOCA in the BWROG report, the NRC technical assistance contractor, Idaho National Engineering & Environmental Laboratory (INEEL), performed a confirmatory calculation for NRC using the SCDAP/RELAPS computer ,
code for thermal-hydraulic calculations and the FRAPCON3 computer code for fuel rod failure l calculations. INEEL prepared a technical evaluation report for NRC," Evaluation of Fuel Pin l Failure Timing in Boiling Water Reactors," dated July 1999 (Reference 3). l l
The staff has also performed an independent and specific confirmatory calculation for GGNS using a series of LOCA analyses to evaluate the minimum time to the onset of fission product l release from perforated fuel rods following a postulated design-basis LOCA. The staff used its TRAC-BF1 best estimate system code with a GGNS input model provided by INEEL and the FRAPCON3 code to estimate the fuel initial conditions. j The purpose of the staff analyses was to confirm that the GE analyses in the BWROG report are conservative.
3.1 Desian-Basis Accident j During promulgation and development of NUREG 1465, the staff conducted a review of current plant final safety analysis reports to identify all design-basis accidents in which the !!censee had identified fuel failure. For all accidents with the potential for release of fission products, the class of accidents that had the shortest time until the first fuel rod failure was the design-basis '
LOCA with complete emergency core cooling system (ECCS) failure. Therefore, the staff concluded that a postulated large break LOCA with complete ECCS failure was a reasonable initiator for modeling the earliest appearance of the fuel gap activity (i.e., minimum time to the l onset of fission product release from a perforated fuel rod).
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3.2 Reactor Fuel Desian The most limiting fuel design for evaluation of earliest fuel rod perforation would include the highest peak linear hest generation rate (PLHGR), the highest stored energy, and the highest internal pressure. To determine the most limiting fuel design, GE evaluated the following different BWR fuel designs that are currently in use: GE8, GE98, GE10, GE11, and GE12, and Siemens 8 x 8 and 9 x 9 fuels to determine the most limiting fuel design. In its evaluation of earliest fuel rod perforation, GE considered two major parameters, PLHGR and fuel rod internal pressure.
Using the PLHGR as an input value, GE evaluated the sensitivity of peak cladding temperature (PCT) to different lattice designs, fuel exposures, and vessel designs. These sensitivity studies were necessary to evaluate the impact of such important parameters as depressurization rate, fill gas pressure, radiation heat transfer, and pin power distribution. On the basis of these sensitivity studies, GE determined that GE11 fuel in a 205-inch inside diameter (ID) vessel with a 28-inch ID recirculation suction line (see Section 3.3) would be the most limiting case. The staff reviewed the sensitivity studies submitted by GE and concurs with GE's conclusion.
3.3 Primary Coolant System Desian The most limiting primary coolant system design for evaluation of earliest fuel rod perforation would be the combination of the smallest reactor pressure vessel (RPV) water inventory with the largest primary coolant break flow rate. The combination of these two parameters willlead to the fastest RPV water inventory depletion resulting in earliest RPV core uncovery. GE used geometry ratios for various BWR designs to select the most limiting primary coolant system design. The BWR design with the smallest ratio (i.e., RPV inventory to break flow rate) will have the earliest core uncovery time.
GE determined that the BWR-4 design with an RPV ID of 205 inches and a recirculation pipe ID of 28 inches would be the most limiting primary coolant system design for evaluating the earliest fuel rod perforation. The staff accepted GE's determination that Vermont Yankee (VY) with a 205-inch RPV ID and a 28-inch recirculation pipe ID has the same limiting primary coolant system design. Therefore, the staff's contractor (INEEL) performed a confirmatory l calculation using VY plant data to check the GE analysis for evaluating the earliest fuel rod
,perforation.
3.4 Accident Seauence Models The thermal and hydraulic design characteristics of the core and nuclear fuel data used by the staff, INEEL, and GE are as follows:
Parameter Vermont Yankee GGNS GE/BWROG (INEEL) (NRC) (GE)
Thermal design output, MWt 1593 3833 1880 RPV diameter, in. 205 251 205 Recirculation pipe diameter, in. 28 22/24 28 Number of fuel assemblies 368 804 484 Fuel configuration 8x8 9x9 9x9 l
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3.4.1_Confirmatorv Analysis by INEEL INEEL converted the VY RELAPS/ MOD 3 input deck to the current version of SCADAP/RELAP5. The RELAPS code calculates the overall reactor con! ant sy5iem thermal l hydraulics and iise 9CDAi' code models the reactor core behavior during a postulated reactor j accident. The calcu;aR .A methodoloyes that used SCDAP/RELAP5 and FRAPCON3 are '
given in detailin Reference 3. The VY reactor core consists of a total of 368 fuel assemblies wi'h thermal design output of 1593 MW1. Each VY fuel assembly is configured in an 8 x 8 fuel rod array with one large centrally locat?d water rod that occupies the space that would otherwise accommodate four fuel rods. utn hmiting core design consists of a total of 484 fuel assemblies with thermal design output of 1880 MWt. Each GE11 fuel assembly is configured in a 9 x 9 fuel rod array with two large centrally located water rods that occupy the space that would otherwise accommodate seven fuel rods.
The INEEL calculated two design-basis LOCA transients for the near beginning of life (BOL) I and at the end of life (EOL) using SCDAP/RELAP5 code with corresponding FRAPCON3 core temperature te determine the minimum time from reactor accident initiation to the first fuel rod !
perforation. The results indicated that the high power and high stored energy conditions near the BOL willlead to an earlier fuel rod failure than high fuel rod pressure at the EOL. INEEL calculated 152 seconds for the minimum time to the onset of fission product release from perforated fuel rods for the conditions at the near BOL. The staff finds that the value of 152 seconds calculatea by INEEL indicates that the 121 seconds calculated by GE in the BWROG report is a conservative value.
I 3.4.2 NRC Staff Confirmatorv Analysis for GGNS The staff obtained a GGNS input deck originally prepared by INEEL. This deck was written for a previous version of TRAC and, therefore,1 ad to be modified to run with the BF1 version of TRAC. All of the plant geometry was retained with the exception of the jet pumps. The staff used a jet pump model, which was extracted from another BWR/6 input deck, and changed the number of jet pumps a correspond to the number at GGNS. The deck consists of 46 i components. The major parameters used in the deck are given in Table 1. The staff used information from the most recent core operating limits report (Reference 4) to determine the fuel type being used at GGNS and used the staff's lattice physics methods to predict radial peaking factors for use in the CHAN components. The part-length fuel rods were not modeled and the internal water channels were modeled with water pins. The axial power distribution is shown in Figure 1 and the remaining kinetics parameters were taken to be the default values.
The deck was tested preceding its use by running it to steady state. The steady-state condition is summarized in Table 2. The core exit quality is slightly higher than the plant design and the feedwater temperature was set lower than design to achieve 10 degrees Kelvin of subcooling at the channel inlet. These differences were not entirely unexpected given the fact that the steam separators were modeled using the simple separator option in the vessel component. These differences are considered acceptable.
The TRAC deck was run for the design-basis recirculation suction line break with no emergency core cooling injection and the fuel was allowed to heat up. This scenario was run for several different linear heat generation rates, recirculation line sizes, and pin power distributions, and used two different rod groupings to evaluate some of the potential sensitivities and to ensure
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that ti analysis maximized the heatup rate. Fuel pin internal pressure was taken to be the value 6t the time of peak reactivity (which was assumed to be the time of peak power) and was predicted using the FRAPCON-3 code. The failt.ro temperature was estimated to be 1000 degrees Kelvir using the methods and data in NUREG-0630 (Reference 5).
The PCT as a function of time is presenied in Figure 2. As shown in the figure, the minimum time to failure is estimated to be 172 seconds, in order to confirm that the model is behaving as expected, the staff also examined the steam dome pressure (Figure 3), the break flows (Figure 4) and confirmed that the ECCS injection flows were zero by examining the major edit. '
4.0 CONCLUSION
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The calculated minimum time to the onset of fission product release from a perforated fuel rod following a postul.ated design-basis accident are as follows:
1 INEEL calculation using VY design: 152 seconds !
NRC calculation for GGNS: 172 seconds l GE caldation in BWROG report: 121 seconds
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On the bas s of this evaluation and findings in the technical evaluation report presented by j INEEL, the staff concludes that the BWROG report," Prediction of the Onset of Fission Gas )
I Release Froni Fuel in Generic BWR," dated July 1996, is acceptable for all currently operating BWRs.
Attachments: 1. Tables 1 and 2 i
- 2. Figures 14 Principal Contributors: Jay Y. Lee Walton Jensen ,
Anthony Ulses Date: September 9,1999 l
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5.0 REFERENCES
(O Letter from W. K. Hughey (Entergy) to USNRC, " Submittal of RWROG Report - Prediction of the Onset of Fission Gas Release From Fuelin Generic BwR: Application of NUREG-1465 Source Term for GGNS Nuclear Station Rebaselining Study," May 6,1997, 1 (2) L. Soffer, et al., " Accident Source Terms for Light-Water Nuclear Power Plants,"
NUREG-1465, U.S. Nuclear Regulatory Commission, February 1995.
(3) D. L. Knudson and R. R. Schultz, " Evaluation of Fuel Pin Failure Timing in Boiling Water Reactors," Idaho National Engineering and Environmental Laboratory, July 30,1999.
(4) Letter from J. J. Hagan (Entergy) to USNRC, " Core Operating Limits Report (GGNS-MS-48.0, Revision 5) for Cycle 9 Submittal," August 13,1997.
(5) D. A. Powers and R. O. Meyer, " Cladding Swelling and Rupture Models for LOCA Analysis," NUREG-0630, U.S. Nuclear Regulatory Commission, April 1980.
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Table 1 i TRAC-BF1 input Deck Description i i
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Component Type Number Notarization VESSEL i 12 Axial,4 Radial,2 Theta CHAN 8 27 Axial,7 or 15 Fuel Rod Groups JETP 2 Standard SEPD 0 Used VESSEL Perfect Separator j
Recire Loops 2 22 Nodes Total Steam Lines 2 10 Ncdes Total Feedwater 2 0 (Used FILL Directly Connected to VESS) l i
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l Table 2
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Grand Gulf Model Steady-State Results Parameter Value Total Power 3833 MW, Core Flow 14,630 kg/sec (1.16x10' lb/hr)
Steam Flow 2000 kg/sec(1.59x107 Ib/hr) l Core Exit Quality 15.7 % j Feedwater Temperature 460 K (368 'F) 9 4
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