ML20148G929
| ML20148G929 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 09/30/1976 |
| From: | YANKEE ATOMIC ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20148G922 | List: |
| References | |
| NUDOCS 8011140259 | |
| Download: ML20148G929 (45) | |
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f O APPENDIX C ADDITIONAL LOCA ANALYSIS WITl! . REVISED UPPER llEMISPilERICAL IIEAD TEMPERATURE September 30, 1976 l 1 Yankee Atomic Elect ric Comp?ny 20 Turnpike Road Westborough, Massachusetts 01581 a o 12 34.o gg 7
l GENERAL The Additional LOCA Analysis presented in Appendix B of the Yankee Nuclear Power Station Core XII Peformance Analysis was submitted with the assumption that the upper hemispherical head temperature is equal to the reactor water inlet tempe rat ure . This assumption is based on the fact that a portion (1% by design) of the relatively cooler inlet water flow is directed through a bypass . gap in the inlet nozzles to cool the upper portion of the vessel head. However, recent operating data gathered from another facility indicates the temperature of the water in the upper head to be warmer than the inlet water by about 60% of the reactor inlet-reactor outict water temperature dif fer-ential. An increase in upper water temperature over that used in the ECCS performance calculations has the ef fect of increasing the calculated peak clad temperature in the event of a Loss-of-Coolant Accident. In lieu of actual plant measurements and in accordance with the Staff's directive of August 12, 1976, additional LOCA calculations have been performed for Yankee Roue Core XII with the conservative assumption of Thot (reactor water outlet temperature) in the upper hemispherical head. These calculations consist of:
- 1) A re analysis of the 0.6 DECLG (identified ag the worst break in Reference 1) ,
at beginning-of-cycle (B0C) conditions; ,
- 2) A re-analysis of the next worst break (1.0 DECLS) at BCC conditions to ;
confirm that the worst break has r.ot shifted, and;
- 3) A re-analysis of the worst break (0.6 DECLG) at the current point ,
in the operating cycle (180 EFPD) . The additional LOCA calculations were conducted using the Exxon ECCS ; evaluation model used in Reference 1 and subsequently approved on June 2, 1976( ). I l
. , RESULTS The following figures illustrate the key parameters for the above mentioned breaks with the assumption of T hot in the upper vessel head: Figures 1A through 1L - 0.6 x DECLG 8.7* kw/ft BOC Exxon fuel Figures 2A through 2L - 1.Q x DECLS 8.7 kw/ft BOC Exxon tuci Figures 3A through 3L - 0.6 x DECLG 10.15 kw/ft 180 EFPD Exxon fuel Tables 1 and 2 compared the important parameters for the breaks. Also compared in these tables are the results for the 0.6 DECLG and 1.0 DFCLS breaks previously analyzed with a Teold upper head temperature. EFFECT OF UPPER HEAD TEMPERATURE ASSUMPTION ON PCT AND LIMITING BREAK SIZE The 1.0 DECLS and 0.6 DECLG break sizes were reanalyzed at BOC conditions to demonstrate the effect the upper head temperature assumption has on peak clad temperature (PCT) and to demonstrate that the previously identified limiting break size (0.6 DECLG) does not shift. The PCT for the 0.6 DECLG break increased by 70 F from 1896 F to 1966 F whereas the PCT for the next limiting break, the 1.0 DECLS, increased only 5 F from 1878 F to 1883 F. Tabic 3 compares these results to the break spectrum reported in Reference 1. Also shown in Table 3 is the temperature increase the remaining points in the break spectrum would have to experience in order to exceed the limiting case (0.6 DECLG). It is highly unlikely that any of the less limiting breaks would 0 undergo the relatively large tergerature increase (100 F-209 F) required. Therefore, it is evident that the most limiting break size (0.6 DECLG) does not change. REVISED ANALYSIS AT CURRENT CYCLE BURNUP The limiting break size (0.6 DECLG) was reanalyzed at 10.15* kw/ft ; and a cycle burnup of 180 EFPD with the assumption of hT ot in the upper vessel head. The LHGR was lowered from the 10.50* kw/ft value presented in f the Core XII burnup study in order to alleviate any prob 1 cms due to potential i rod bursting. The revised peak clad temperature for this point vas calculated , to be 1988 F compared to the PCT of 2034 F calculated previously for this , burnup which was performed at 10.5 kw/ft.
- Total linear heat generation rate ,
i (LilGR) __
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. REVISED LHGR VS. BURNUP PERFORMANCE CURVE A revised limiting linear heat rate vs. burnup curve is shown in [
Figure 4. The points at 10 EFPD,180 EFPD and 389 EFPD for both Exxon f I and Gulf fuel have been lowered ' proportionately in accordance with the results of the analysis for the 0.6 DECLC at 10.15 kw/ft and 180 EFPD burnup. The initial point for Exxon fuel remain unchanged due to the .' acceptable results demonstrated by the 0.6 DECLG That analysis at BOC. , REFERENCES
- 1) Proposed Change No. 125, Supplement 8, (February 20, 1976).
Amendment No. 26 to Facility Operating License Ac DPR-3, June 2, 1976. 2) l I l i l l
TABLE 1 . YMiKEE ROWE CORE XII LOSS OF COOLA'iT ACCIDENT UPPER VESSEL llEAD TEMPERATURE STUDY SUFBfARY OF RESULTS(1) 1.0 DECLS 0.6 DECLG 0.6 DECLC Parameter . Total Feat Generation Rate, kw/ft 8.7 8.7 8.7 8.7 10.5 10.15 Rod Linear lient Generation Rate, kw/ft 8.47 8.47 8.47 8.47 10.22 9.88 515 560 515 560 515 560 Upper licad Temperature, F ' F 1878 1883 1896 1965 2034 1988 Peak Clad Temperature, 4.04 4.04 4.04 4.04 4.04 4.04 Peak Clad Temperature Location, Ft. 1.69 Maximum Local R Z /ll 1.44 1.19 1.53 1.50 2.12 2 0 neaction, % 4.04 4.04 4.04 4.04 4.04
!!aximum Local ZR /II20 Reaction Location, Ft. 4.04
<1 <1 <1 <1 <1 <1 ,
Total Core Z R/II 2 0 Reaction, I 0.0 0.0 0.0 0.0 180. 180. Burnup, EFPD (1) Calculations perforced at the following conditions: License Core Power lht 600 Power Used for analysis, Mwt 618 . Accumulator Water Volume, Ft 700 Fuel Type ENC e., t 6 1
-A 1-
TABLE 2 . VANKEE ROWE CORE XII LOSS' OF COOLANT ACCIDENT - UPPER llEAD TEMPERATURE STUDY TIME SEQUENCE OF EVENTS Event Time, Seconds Break Size: 1.0 DECLS 0.6 DECLG Burnup/LilGR B0C/8.7 kw/ft B0C/8.7 kw/ft Event Upper IIcad Temperature: 515 F 5600F 5150F 5600F, Pipe Rupture 0 0 0 0 Begin Accumulator Spillage 0 0 0 0 Loss of Offsite Power 0 0 0 0 Safety Injection Signal 7.54 7.58 7.54 7.58 Accumulator Injection, Intact Loops 19.10 18.28 19.88 19.42 Safety Injection Pump Flow Start 32.54 32.58 32.54 32.58 End of Blowdown (E0B) 32.50 30.95 34.73 33.70 End of Bypass (E0BY) -39.50 33.56 40.30 -39.34 Bottom of Core Recovery (BOCREC) 102.0 100.65 102.49 101.60 Accumulator Empty 108.3 107.80 109.79 109.30 Peak Clad Temperature Reached 112.8 114.96 117.10 116.34
+
a Table 3 . Yankee Rowe Core XII Loss of Coolant Accident Upper Vessel llend Temperature Study Comparison of Results to PCT's Reported In Reference 1 Peak Clad AT Required to Temperature, F Exceed Limiting Case, OF LHGR kw/ft: 8.7 8.7 Burnup, EFPD: BOC EOC ' Upper llead Temperature, F: 515 515 Break Size 0.6 DECLG 1896 1966 -- 1.0 DECLS 1878 1883 -- 1866
- 100 0.6 DECLS 1861
- 105 1.0 DECLG 1817
- 149 0.4 DECLG ,
1757
- 209 0.4 DECLS
- Not re-analyzed b
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~ - 50-29 NRC DISTRIOUTION rOn PART 50 DOCKET MATERIAL To: FROM: YANKEE ATOMIC ELECTRIC CO O^tc Or DoCuMCNT USNRC WESTIOROUGH, MASS 10-8-76 L H HEIDER DATE ACCEIVED 10-8-76 to rica ONOTOnizt0 rnOP INPUTFORM NUMDER OF COPIES RECElVED oniGIN AL NCLASSIFIED # ' COPY /g
- DF S ,* H I P TION ENCLOSURE Q LTR N0TORIZED 10-6-76 REF THEIR LTR 8-27-76 ADDITIONAL CORE XII I.OCA ANALYSIS WITH LTR TRANS THE IT>LLCWING. . . . . . . . . . . . ELEVATED UPPER HEMISPHERIJAI, HEAD TEMPERATURE WITH ATTACHEJ FIGURES. . . . . . . . . .
ACKNOWLEDGED 00 M0TREMnyE \ PLANT NAME: YANKEE R0WE...... . e I a
- -. SAFETY FOR ACTIOf1/INFORMATiON Erp[IL_ / O - /C - 74 /2/'/2-
! I ASSIGNED AD: I ASSI0lg.Q AQ: M.hrsich t.,nic.F: 6C H-leEO MIL snAnu1LCn1Er e ! FROJECT MANAGER: PROJFST lk%'AGER: / LIC. ASST.: GhC/ par ' h LIL._ ASST.: L_ l-j $__ m_ INTERNAL DISTRIBUTION WEG FILEj SYSTEM _S SAFETY PLANTJYS.TEMS SITE _SAEETY_b // I _ & __ E (,p) NRC FDR llEINEMAN TEDESCO __El?V_IE0JNALYSIS SCHP,0EDER BENAROYA _pyNTON & MULLER h2i4D LAINAS ., _ COSSICK & STAFF ENGINEERING IPPOLITO Eb3'IEQ_TECIL. i MI_PC P MACCARRY KIRKWOOD FRRST CASE ' KNIGHT BAIJ ARD dliANAUER SINWEIL OPERALING REACTORS _SEANCLER JARl.,FtSS PAWLICKI STELLO SI'"E TFCH ; PROJECT _PANAGEMENT REACTOR SAFETY OPERATING TECL___ GAMMILL 30YD ROSS / _B_ISEhM[r SIEPP L P, COLLINS NOVAK / _.JRAO __HUI2/dl , llCUSTON ROSZTOCZY / BAEPR 'Y PETERS _0N CHECK / BUTLER I __SlIE_ABALYSIS MELTZ _. / GRTMES VOLLMER \ i_.EE.LYEMES Al & I BURCH SKOV110LT SALTZMAN C_J._ COLLINS RIITBERG ynrCrp , s EXTERNAL DISTRIBUTION COfM ROL NUMBER I ,' M PDR: M4GM-t .LTL _G M ASL N/.T ' LAB : E R00KilAVEIL.NAT., LAD _ ' ~ REG. VIE ULRIKSON(ORNL) ,O LA PDR =. M] HIC: ' SIC - ASLn: lACRSl~bCYS Il0LDING41Ft) CONSULTANTS + } Q k{ f ~ ~ - - rr}}