ML19319B966
| ML19319B966 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/08/1978 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML19319B964 | List: |
| References | |
| BAW-1489, NUDOCS 8001290699 | |
| Download: ML19319B966 (16) | |
Text
{{#Wiki_filter:.- - (, . L-] Docket No. 50-346 Operating License NPF-3 Serial No. 443 June 8, 1978 BAM-1489 May, 1978 Revision 2, 6/8/78 ATTACHMENT 1 TO APPLICATION TO A'!END OPERATING LICE:ISE FOR REMOVAL OF BURNA3LE POISON ROD ASSEMBLIES
- Davis-3 esse Nuclear Generating Station, Unit 1-(This is to be added at the end of BNJ-1439 with Rev. 1 included).
l l l l 3ABC0CK & UILCO:; Power Generation Group Nuclear P.Power Generation DivisioB 0 01290 6 - O. Box 1260 Lynchburg, Virginia 24505 revised 6/8/78
TOLEDO ggERy w eoisom Docket No. 50-346 LOWELL E. RCE
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License No. SPF-3 14m m-naz June 8, 1978 Serial No. 443 Director of Nuclear Reactor Regulation Attention: Str. John F. Stolz, Chief Light Water Reactors Branch No. 1 , Division of Project 'ianagement United States Nuclear Regulatory Commission Washington, D. C. 20555
Dear Str. Stolz:
As a result of recent telephone conversations with your Licensing Project Manager (Mr. Leon Engle) a- your review staf f, we are revising part of the Accident and Transient analysis of Revision I to " Attach =ent 1 to Application to Amend Operating License for 3urnable Poison Rod Assemblies and Orifice Rod Assemblies," which was attached with our letter to you dated ?!ay 26,1978 (Serial No. 439) . The revised pages are given in R: vision 2, which is attached. Also attached is a list of Power Ascension Tests which will be perfor"Jed in addition to the testing already committed to. Yours very truly, f ' I / Attachments jh a/12 h t l THE iCLECO ECISCN COMPANY ECISCN PLAZA 3CC MACISCN AVENUE TCLECO, CHIO 43552 g[ l 1 l l
4 Contents for Revision 2 to BAW-1489 TEXT PAGE Rod Withdrawal from Power . . . . . . . . . . . . . . . . . . . . B-1 Thermal Hydraulic Evaluation of Anticipated Transients . . . . . . B-2 . Loss of Coolant Flow Transients . . . . . . . . . . . . . . . . .. B-2 Excess Heat Removal Transient . . . . . . . . . . . . . . . . . . . B-3 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-4 TABLE Table 1 Summary of Minimum DNBR Results for Limiting Transients . . 3-4 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-5 FIGUItES Figure 15. 2.2-2 o f Davis-Besse 1 FSAR. . . . . . . . . . . . . . . 3-6 Figure 15.2.2-2A . . . . . . . . . . . . . . . . . . . . . . . . . B-7 Figure 15.2.2-23 . . . . . . . . . . . . . . . . . . . . . . . . . 3-8 Figure 2.1-1 (Revised 5/16/78) of Technical Specifications . . . . 3-9 Figure 1 - Minimum DNBR and Maximum Fuel and Cladding Temperature versus Time for 4 PCD . . . . . . . . . . . . . . . . 3-10 Figure 2 - Film and Gap Coefficients versus Time for 4 PCD . . . . 3-11
- Figure 3 - Power and Flow Transients for 1 PCD . . . . . . . . . . 3-12 Figure 4 - MDNBR versus Time for 1 PCD, Flux to Flow Analysis. . . 3-13 Figure 5 - Excessive Heat Removal. . . . . . . . . . . . . . . . . 3-14 r
3-1 Revised 6/8/78 l l 1
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ROD WI'IEDRAWAL FROM POWER The rod withdrawal event has been examined with respect to the potential for DNB occurrence during this event. The basis for our evaluation was the proximity of the main parameters influencing DNB during the transient (i.e., core outlet pressure, core outlet temperature, and thermal power) to the steady state DNB safety limit for 112% power at design peaking and 2/2 RC pump operation. This DNB safety limit is defined in Figure 2.1-1 of the DB-1 Technical Specifications (as revised in BAW-1489, Rev. 1 - See Figure 7-la). As shown in Figure 15.2.2-2 of the DB-1 FSAR, for the spectrum of anticipated reactivity insertion rates, system pressures are always increasing for the rod withdrawal event such that peak core outlet pressures are always greater than the initial condition of 2185 psig. Similar curves for peak core outlet temperacare and thermal power are shown in attached Figures 15.2.2-2A and 15.2.2-2B. These curves confirm that core outlet temperatures are always less than 617F and thermal power always less than 112%. When these limiting conditions are superimposed on the Figure 2.1-1, also attached, it is clearly shown that transient conditions produced by the rod withdrawal transient are always maintained by the reactor protection system within the range of acceptable operation without DNB occurrence. Due to the known margin to DNB as established in the results presented above, B5W was not required in the licensing of the DB-1 FSAR to provide detailed hot channel analysis for the rod withdrawal event. At the request of the NRC, B&W has reviewed the rod withdrawal event with respect to whether it represents the limiting case for DNB among the entirety of the normal operating occurrences. B&W han recently submitted (Reference Consumers FSAR and BSAR), for similar plant designs, detailed hot channel evaluations using the B&W2 CHF correlation. The evaluations encompassed a wide power range and it was concluded that the rod withdrawal event is not the lLmiting DNB transient. To illustrate this conclusion, results are shown below which establish that the four pump coastdown would always be expected to be a more limiting DNB transient than the rod withdrawal event: MINDtDi DNBR (BAU-2) 177FA 205FA W-B Core Mk-C Core 2452 MWt 3800 MWt 4 Pump Coastdown 2.0 1.4 Rod Withdrawal 2.21 1.595 l l l l P-1 Revised 6/8/78
THER&tL HYDRAULIC EVALUATION OF ANTICIPATED TRANSIENTS An evaluation was performed to determine the effects of the removal of burnable poison rod assemblies (BPRA's) and orifice rod assemblies (ORA's) from the Davis-Besse 1 reactor. This evaluation considered transients which have been shown, in the FSAR (1) and Fuel Densification Report (2), to be more limiting than other anticipated transients (in terms of DNBR). Conditions analyzed include loss-of-coolant-flow and excessive heat removal transients. LOSS OF COOLANT FLON TRANSIENTS All loss-of-coolant-flow (LOCF) transients, with the exception of one pump coastdown from four pump operation, will result in a reactor trip which is initiated by pump power monitors. Therefore, the most limiting LOCF transient for which the pump power monitors provide DNB protection is the four-punp coastdown. For the one pump coastdown (from four pump initial operation), reactor trip is initiated by the flux / flow trip function. This transient, which is the basis for the determination of the flux / flow trip setpoint for Davis-Besse 1, is more limiting than the four pump coastdown by virtue of the flux / flow trip setpoint specified in the Technical Specifications. For this evaluation, the , ef fects of the removal of BPRA's and ORA's from the Davis-Besse 1 core have been incorporated by increasing the core bypass flow from 6.04%, as used in References 1 and 2, to 10.75%. In addition, the minimum reactor coolant system flow rate has been increased from 352,000 cpy to 387,200 GPM. As a result, LOCF transients demonstrate improved safety margins relative to analyses previously performed for this core. Table 1 summarizes the results of these analyses. The four pump coastdown analysis performed for the revised core configuration resulted in a calculated minimum DNBR of 1.85 (BAW-2). Similar analyses per-formed for the FSAR and Fuel Densification Report (references 1 and 2) resulted in calculated minimum DNBR values of 1.49 (U-3) and 1.82, respectively. Thus, the revised core configuration combined with the increased system flow basis results in improved thermal margin. Significant results for this analysis are shown in Figures 1 and 2. The transient input is shown in Figure 3.4-3 of Reference 2. The one pump coastdown analysis, as discussed above, forms the basis for the determination of the flux / flow trip setpoint. For this evaluation the existing flux / flow trip setpoint, as specified in the Davis-Besse 1 Technical Specifications (1.08), was evaluated to determine the available thermal margin. The procedure used was to first determine the flow fraction associated with the flux / flow setpoint (corrected for flow measurement error and flow signal noise) and 102% indicated power. The flow fraction was then related to the appropriate pump coastdown curve (one pump for Davis-Besse) to obtain the time at which the trip signal would occur. The delay from trip signal to start of control rod motion (1.5 seconds) was then added to determine the time at which power would start to decrease as a result of reactor trip. The resulting power transient and the flow coastdown transient (Figure 3) were then input to the RADAR (3) code, and a transient DNBR calculated, based upon an assumed initial reactor power level of 108%. The initial DNBR and hot channel conditions used in the RADAR code were determined from core thermal hydraulic analyses performed with the CHATA (4) and TEMP (5) codes consistent with References 1 and 2 . The result of this analysis, shown in Figure 4 and summarized in Table 1, is a minimum DNBR of 1.45 (BAW-2), which provides 12% margin to the 1.30 CHF correlation limit. This margin is sufficient to offset a DNBR penalty resulting from fuel rod bow equivalent to a fuel burnup of 33,000 MWD />nt or greater. B-2 Revised 6/8/73
4 EXCESSIVE HEAT REMOVAL TRANSIENT The Davis-Besse 1 FSAR indicates that a limiting transient, in terms of DNBR, is the reduction in feedwater temperature. This transient has been reanalyzed for the revised core configuration and the results are shown in Figure 5 and summarized in Table 1. The transient inouts are the same as shown in Figure 15.2.10-1 o f the FS AR. The minimum DNBR is 1.80 (BAW-2), versus 1.43 (W-3) reported in the FSAR. Thus, it is concluded that core safety margins are in excess of those reported in the FSAR. S12! MARY The analysis results presented herein demonstrate that the revised Davis-Besse 1 core configuration, when combined with the increased reactor coolant system flowrate, result in improved core safety margins relative to those defined in References 1 and 2. The increased flowrate used for these analyses (110% of design flow) is supported by the measured flowrate value of 113. 5% of design flow. Therefore, it is concluded that the removal of BPRA's and ORA's from the Davis-Besse 1 core presents no safety problems. B-3 Revised 6/8/78
TABLE 1 Summary of Minimum DNBR Results for Limiting Transients Transient FSAR(1) Fuel Revised i Densification Cycle 1 Report (2) Configuration (tJ-3) (BAW-2) (BAW-2) One Pump Coastdown (flux / flow trip) N.R. 1.37 (N.R.) 1.45 Four Pump Coastdown (pump monitor trip) 1.49 1.82 1.85 Excessive Heat Removal 1.43 N.R. 1.80 N.R. - not reported i l B-4 Revised 6/8/78
REFERENCES
- 1. Davis-Besse Unti 1, Final Safety Analysis Report.
- 2. Davis-Besse Unit 2 Fuel Densification Report, BAW-1401, Babcock & Wilcox, April, 1975.
- 3. RADAR - Reactor Thermal and Hydraulic Analysis During Reactor Flow '
Coastdown, BAW-10069, July, 1973.
- 4. CHATA - Core Hydraulics and Thermal Analysis Program, BAW-10110,,Rev. 1, May, 1977.
- 5. TEMP - Thermal Enthalpy Mixing Program, BAW-10021, April,1970.
B-5 Revised 6/8/78
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