ML20002A520
| ML20002A520 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 01/10/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20002A517 | List: |
| References | |
| NUDOCS 8011170463 | |
| Download: ML20002A520 (10) | |
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Ii SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 1
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SUPPORTING AMENDMENT NO. 34 TO FACILITY OPERATING' LICENSE NO. DPR-3 u
YANKEE ATOMIC ELECTRIC COMPANY I
YANKEE NUCLEAR POWER STATION (YANKEE-ROWE)
DOCKET NO. 50-29 i
Introduction l
By application Qted October 8,1976, Yankee Atomic Electric Company 1
(the licensee) proposed changes to the Technical Specifications, appended to License No. DPR-3 for the Yankee-Rowe reactor. The proporal involves a revision of the " Core XII Allowable Peak Rod LHGR versus l
Exposure" specified in' Figure 3.2-1 to reflect changes of the allowable LHGR values based on the revised ECCS performance analysis submitted i
I with the October 8,1976, application and the additional information l
provided in the licensee's letter dated December 27, 1976, i
Dis cussion l
i Yankee-Rowe has been operated within the restrictions on the allowable peak rod LHGR imposed by the NRC's Order for Modification of License issued on August 27, 1976.
Specifically, the Order modified the Technical Specification limit by reducing the allowable peak rod LHGR by 0.85 kw/ft. This provided assurance that ECCS performance at the 2
facility conforced to all criteria set forth in 10 CFR 550.46(b),
despite the assumption in the previous ECCS performance analyses that the reactor vessel upper head temperature is equal to the reactor inlet water temperature. The Order also required that the licensee provide
+
i as soon as possible revised calculations, using an approved evaluation model, with correct input for upper head water temperature, or assuming i
that the upper %ad water temperature equals reactor vessel outlet tempe rature.
On )ctober 8,1976, the licensee submitted the revised l
ECCS' analysis in compliance with the Order.
On December 27, 1976, i
the licensee provided additional information as a result of discussions l
with the staff.
I l
Eval uati on-g Previously, ECCS performance analyses for all Westinghouse reactors, was at the same~ temperature as the reactor inlet water (pper head region including Yankee-Rowe, assumed that the reactor vessel u j
coldleg). This j-assumption was based on the existence of a bypass flow path with a small-percentage of the flow routed directly to the upper head region.
Recent
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, t operating data from another facility indicated that the upper head water temperature was higher than the reactor inlet water temperature by about 60% of the temperature difference between inlet and outlet.
An increase in upper head region water temperature increases the cal-culated peak clad temperature in the event of a LOCA.,
4 In lieu of. actual plant measurements revised LOCA analyses were required with upper head temperature equal to reactor outlet temperature to establish new operating limits for Yankee-Rowe.
Complying with the NRC's Order for Modification of License, the licensee initially performed additional. LOCA calculations consisting of:
(1)' A reanalysis of the 0.6 DECLG (identified as the -
it break
~
in previous analysis at beginning-of-cycle (BOC'
,.idi tions ;
(2) A reanalysis of the next worst break (1.o DECLS) at B0C conditions; (3) A reanalysis of the worst break (0.6 DECLG) at the current point in the operating cycle (180 EFPD).
The calculations were done with the Exxon ECCS evaluation model approved for Yankee-Rowe and used in previous ECCS performance analyses.
Tables 1 and 2 compare the important parameters for these breaks. Also compared in the tables are the results for the 0.6 DECLG and 1.0 DECLS.
breaks previously analyzed with a Tcold uppe'rlead temperature.
It should
~
be noted that the Peak Clad Temperature (PCT) for the 0.6 DECLG break increased at B0C by 700F from 18960F to 19660F whe;eas the PCT for the next limiting break, the 1.0 DECLS, increased only 50F from 1878 F to 0
1883 F.
Table 3 compares these results to the break spectrum analyzed previously for Core XII. Also shown in Table 3 are the temperature increases the remaining points in the break. spectrum would have to experience in order to exceed the limiting case (0.6 DECLG). The licensee concluded that it is unlikely that any of the less limiting breaks would undergo the relatively large temperature increase (1000 -2090) required to become F
limiting and therefore contended that the most limiting break size (0.6 DECLG) does not change.
The limiting break size (0.6 DECLG) was reanalyzed at a LHGR of 10.15 kw/ft T ot in the upper and a cycle burnup of 180 EFPD with the assumption of h
vessel head. The LHGR was lowered from the'10.50 kw/ft value used in the Core XII analysis because the Exxon Fuel in Core XII exhibits a rod burst threshold slightly above-20000F and the analysis at a LHGR of 10.5 kw/ft resulted in a predicted PCT of 20340F At 10.15 kw/ft the PCT was calculated to be 19880F.
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h The staff did not agree that the licensee had shown conclusively that the 0.6 DECLG break was still the limiting break for the changed assumption. The staff position was based in part on our experience with another plant where a reanalysis with the Exxon evaluation model did show that the limiting break size changed from 0.6 DECLG to 0.8 DECLG as a Tcold to T ot with a-h result of changing the upper head temperature from substantial increase in PCT becayse for Yankee-Rowe the difference in break area between CD = 0.6 and 'D = 1.0 is relatively small since the inlet pipe has a cross sectional area of only 1.42 square feet.
In addition the variation of PCT was previously calculated to be strail as C,is varied from 0.4 to 1.0 so that a large variation in PCT D
between CD of 0.6 and 1.0 would not be expected.
However it was noted that the calculated peak clad temperature at CD = 0.6 FCLG was 19880F, just 120F below the predicted rod burst threshold for the Exxon fuel.
Staff experience with the Exxon Evaluation Model indicates that guillotine type breaks result in a maximum PCT between CD = 0.4 and UD = 1.0 but ghat for split or slot type breaks the PCT increases with break size and D = 1.0 generally results in the maximum PCT. We therefore concluded that the reanalysis of the 1.0 DECLS break was sufficient to show that-the slot breaks are not limiting.
The licensee was requested to perform additional analyses to verify that the limiting break size had been identified. These analyses were performed by the licensee and submitted on December 27, 1976. They CD=0.4, C =0.6 and 'D=0.8 guillotine type ccasisted of three cases:
D breaks.
The analyses were carried out to the beginning of the reflood period (80CREC).
These analyses also included the POST-CHF return to nucleate boiling lockout heat transfer correlation model which is described in Exxon Nuclear Company's technical report XN-76-44, September 1976. This was d:ne in response to a specific request in our letter to the licensee dated December 10, 1976.
The staff agreed to consider calculations made out to BOCREC as sufficient to indicate the limiting break size for Yankee-Rowe for the following reasons:
(1) The.reflood transient for Yankee-Rowe is short compared to most other reactors, on the average of 15 seconds from BOCREC until PCT occurs.
(2) The containment ~ backpressure which is a significant parameter during reflood is assumed to be constant and is selected to be conservative I
with respect to all blowdown cases considered.
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TABLE 1 j
YANKEE ROWE CORE XII LOSS OF COOLA!rr ACCIDENT UPPER VESSEL llEAD TEMPERATURF. STUDY SUSDIA!Y OF RESUllrS(I)
Pa rame ter 1.0 DECLS 0.6 DECLC 0.6 DECLC 8.7 8.7 8.7 8.7 10.5 10.15 Total IIcat Cencration Race, kw/f t
~
8.47 8.47 8.47 8.47 10.22 9.tf 8 Rod Lincar llcat Generation Rate, kw/ft Upper Head Temperature, "F 515 560 515 560 515 560 Peak Clad Temperature, F
1878 1883 1896-1966 -
2034 1938 h
Peak Cla,1 Temperature Location, Ft.,
4.04 4.04 4.04 4.04 4.04 4.04 2
1.44 1.19 1.53 1.50 2.12 1,69 Maximun Local Za/ti 0 Reaction, %
Haximum Local Zg/l!20 Reaction Location, Ft.
4.04 4.04 4.04 4.04 4.04 4.04 2
_ <1
<1
<1
<1
<1
<1 Total Core Zg/I! 0 Reaction, %
0.0 0.0 0.0 0.0 180.
180.
Burnup, EFPD (1)
Calculations perfor: red at the following conditions:
License Core Poucr Itwt 600 Crrq Power Used for analysis, Hwt 618 td. Accu. ulator Water Volume, Ft 700 fFuelType ENC ^1&u WJ E2* t c r )' O e @5) DES., mm Yi cae.S wa b ENd )
_m u$ '1 - ' M"* "~ ++e~ k-u F'n ii 4 '" i TABLE 2 YANKEE ROUE CO!!E XII I.OSS' 0F C001. ANT ACCIDENT UPPElt IIEAD TEMPEl ATURE STUDY + TIMF. SEQUENCE OF EVENTS Event Time, Seconds Break Size: 1.0 DECLS 0.6 DECLC ~ Burnup/LilCR BOC/8.7 kw/ft BOC/8.7 kw/ft. Event Upper licad Temperature: 515"F 5600F 5150F 5600F - Pipe Rupturc 0-0 0 0 Segin Accumulator Spillage 0 0 0 0 Loss of Offsite Power '9 0 0 0 Safety Injection Signal 7.54 7.58 7.54 7.58- - Accumulator Injection, Intact I. cops 19.10 18.28 19.88 19.42 b. Sarcty Injection Pump Flow Start 32.54 32.58 32.54 32.58 End of B1c,wdown ~(E03) 32.50 30.95 34.73 33.70 .End of Bypass (20BY) 39.50 38.56 40.20 - 39.34 tottom 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 T
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Table 3 e Yankee ~ Rowe Core XII Loss of Coolant Accident Upper Vessel licad Temperature Study Comparinon of Res_ults With Previous LOCA Analysis t I' Peak Clad AT Required to _ Temperature "F-Exceed Limiting Case, F LilGR kw/ft: 8.7 8.7 p. Burnup, EFPD: BOC BOC i Upper llend Temperature, F: 515 560 Break Size 0.6 DECI.0 1896 1966 g 1.0 DECLS 1878 1883 U.6 DECI,S 1866 100 1.0 DECLG 1861 105 O.4 DECI.G 1817 149 p' O.4 DECLS 1757 209
- Nc re-analyzed
k 2-- w- ~ - - - - - - - G (3) Hot rod conditions, i.e. hot rod clad temperature and stored energy input to the reflood and fuel heatup calculations were calculated to beginning of reflood and were compared at that point to determine which case would result in the highest PCT. The 0.6 DECLG case was clearly limiting with respect to both PCT and fuel pellet average temperature of the peak power rod at End of . Bypass (E0BY). Table 4 presents the results of the reanalyses. A comparison is also made of the previous analysis without the returg to nucleate boiling lockout but with the upper head temperature at ' hot. The staff concludes that the analyses described above adequately determine C =0.6 DECLG. D that the limitino break size remains The limiting break, C =0.6 DECLG was analyzed through the reflood and D heatup period and the results are presented in Table 5. Table 5 also compares the results of the previous analysis without return to nucleate boiling lockout. Revised curves showing allowable peak rod LHGR (Figure 3.2-1) as a function, of burnup were submitted with the analyses. The curves for the Exxon fuel which is limiting and the Gulf fuel which had been previously burned in Cycle XI were both lowered proportionally based on the calculated reduction in LHGR for the Exxon fuel at 180 EFPD. The staff agrees that the proportional lowering of both curves based on the limiting analysis at 180 EFPD on the Exxon burnup curve is acceptable since the Exxon fuel is limiting, the LHGR curve is well predictable from 180 EFPD to end of cycle, and sufficient margin exists between limiting LHGR and clad damage threshold. The staff finds that the licensee has provided acceptable revised ECCS calculations, using an approved evaluation model, with the. assumption that the upper head water temperature equals the reactor vessel outlet temperature. The l'. 2nsee has also included an acceptable correction in the evaluation nodel that precludes the use of a nucleate boiling heat transfer correlation during blowdown after CHF has been predicted. The staff also finds that the results of the revised ECCS calculations verify that ECCS performance at Yankee-Rowe will conform to the criteria set forth in 10 CFR 550.46(b) for operation with Core XII within the revised allowable peak rod LHGR as proposed in the licensee's October 8, 1976, application. Accordingly, the proposed revised Figure 3.2-1 is acceptable for incorporation in the Technical Specifications concurrent with the termination of NRC's August 27, 1976, Order for Modification of License, c t l
f Table 4 Yankee Rowe Core 12 LOCA Analysis Results of Additional f 0CA' Analysis with Elevated Upper-Head Temperature (UHT) and Post-CHF Return to Nucleate Boiling ~ Lockout (RNBLO) Analysis with Elevated UHT Analysis with and RNBLO Elevated UHT Only Q Break Size: 0.8 0.6 0.4 0.6 ? Break Type: .DECLG DECLG DECLG DECLG Cycle Conditions: BOC BOC B0C BOC. Total Max. LHGR, kw/ft: 8.7 8.7 8.7 8.7 cn E0BY, seconds: 38.30 39.54 43.54 39.54 PCT Rod Clad Temperature at E0BY, UF 1243.1 1317.1 1262.1 1314.90 BOCREC, seconds 100.46 101.64 105.14 101.64 PCT Rod Clad Temperature at B0CREC, OF 1751.0 1824.5 1765.4 1823.0 Pellet Average Temperature of Peak 1425 1492 1425 1490.2 i Power Rod at E0BY 64 9 "Y- -. -m m - m
, ~ v.. l Table 5 ~ d -. Yankee Rowe Core 12 Limiting LOCA Analysis-With Post-CliF Return-to-Nucleate Boiling Lockout (RNBLO) T ot h - Ar.d Upper IIcad Temperature 0.6 DECLG Break Size: Maximum Total tilGR: 10.15 kw/f t 18.0 EFPD 4 Fuel Exposure: Exxon Fuel Type: Without RNBLO With RNBL Parameg 39.54 39.54 End-of-Bypass, seconds' 1282.1 F 1280.7 PCT Rod Clad Temperature 0 E0BY, O 101.6. 1 01.6 Bottom of Core Recovery, seconds-1875.3 PCT Rod Clad Temperature 0 B0CREC, F 1872.0 112.1 111.7-Time of Peak Clad Temperature, seconds 1985.1 1987.5, Peak Clad Temperature, OF n r t i
g .. ~ h 10 - We have determined that'the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental. impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant-from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment. Conclusion We have concluded, based on the considerations discussed above, that: (1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the C6mmission's regulations and the issuance' of this amendment will not be inimical to the common defense and security or to the health and safety of the public. Date: January 10, 1977 4}}