ML20148G055

Safety Evaluation Supporting Confirmatory ECCS Small Break Analysis Acceptance Amend 25 to License DPR-3

ML20148G055
Person / Time
Site:	Yankee Rowe
Issue date:	10/07/1976
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20148G041	List: ML20148G037 ML20148G043 ML20148G048 ML20148G055 ML20148G060
References
NUDOCS 8011070150
Download: ML20148G055 (4)

Text

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING ACCEPTANCE OF CONFIRMATORY ECCS SMALL BREAK ANALYSIS i

AMENDMENT NO. 25 TO FACILITY OPERATING LICENSE NO. DPR-3 YANKEE NUCLEAR POWER STATION (YANKEE-ROWE) $

i DOCKET NO. 50-29 ,

Introduction 1 By application dated May 11, 1976, and supplements dated May 11, 17, 19, 1976, and June 16, 1976, Yankee Atomic Electric Company (the licensee) ,

oroposed a change to the Technical Specifications appended to License No. 2 D?R-3 for the Yankee-Rowe reactor. The proposal involved a change of the '

set pressure for the safety injection accumulator, specified in Section  ;

3.2.e(4), from 410 psig to 337 psig 10 psig. The staff's safety evaluation, .

de ed May 19, 1976, concluded that adjusting the accumulator pressure to 1 a:ch previous ECCS calculations was acceptable for large break LOCAs.

Although the analyses submitted in May 1976 were adequate to demonstrate

' ht the revised accumulator pressure resulted in ECCS performance in

. qy conformance with the Commission's Acceptance Criteria for large breaks 5 ar.d for small breaks of 4-inch diameter or less, these initial analyses did not provide computations for small breaks larger than 4-inch diameter. y$ pt The licensee asserted that the larger breaks (e.g. 5-inches) would result in faster depressurization and more rapid injection of emergency core %n

colant which would result in more rapid recovering of the core. The 4 Staff's May 19, 1976 safety evaluation concluded that the 5-inch break is

'.icely to be less limiting than the 4-inch break. The Staff accepted the g

Q li:ensee's correction of accumulator pressure on an interim bases, provided Mr that the licensee submitted improved small break LOCA analysis to confirm that the most limiting small breaks had been identified and conformed $g to the Commission's Acceptance Criteria. My The licensee submitted such analyses on June 16, 1976. This Safety Evaluation h l describes the Staff's evaluation of this confinnatory analysis. &(

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Staff Evaluation orior to the June 16, 1976 submittal, small break analyses were performed y i

by the licensee using the WFLASH code and submitted to the NRC on July 31, Qk I

1974. These analyses considered 2, 3 and 4-inch diameter breaks. Results of these analyses indicated that the core never became uncovered for the k

Jih 2" break. A peak clad temperature (PCT) of 7500F was calculated for the n.1 3-inch break and the 4-inch break was determined to be limiting with a PCT of 13000F occurring 235 seconds after break initiation, W

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The 4-inch break was re-analyzed using the ENC WREM based evaluation model, Results of this analysis were included in the licensee's October 10, 1975 g(

application for a license amendment. The calculated PCT during the period g

of core recovery was determined to be 1400 F at 230 seconds. However, a 4

U fff PCT of about 1850 F was calculated to occur at about 9 seconds due to a $

momentary flow stagnation. Small breaks larger than 4-inch diameter had Ki not been previously conducted using the SLAP code in 1971 which indicates that the core never uncovered.

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The licensee's confirmatory small break analyses were conducted with the f$$

M approved RELAP-4-EM SMALL BREAK code (RELAP4-EM/00311/07/75 95 ENC 20). .r For fuel heatup calculations, the approved T00DEE2 ENC 13 Code was used.

Small breaks of 2.00, 2.25, 4.0,5.0 and 10 inch ID were analyzed.

Corrected accumulator pressure and flow resistances were used. Direct h spillage of ECCS fluid from the safety injection line in the broken loop  ;;

to the containment floor was assumed to occur for the 2.25 inch ID and "Y all larger breaks. The 2.25 inch ID break corresponds to a complete -C severence of a safety injection line. No direct spillage of ECCS fluid [tg to the containment was assumed to occur for breaks smaller than 2.25- ru inches ID. In these cases, the reactor coolant system (RCS) pressure is felt in each loop of the ECCS system including both pumps and the $ M accumulator. A The peak clad temperature analysis was performed using the T00DEE2 y

digital computer code. For the 4, 5 and 10-inch ID breaks, the peak X power fuel rod was divided into 11 axial nodes. Initial fi el rod &n temperature distributions were obtained from a steady-state analysis using a RELAP4-EM/ HOT CHANNEL model with a corresponding axial nodalization.

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Time dependent fluid conditions required as T00DEE2 inpi . came from the blowdown results.

M.0 For the 2.0 and 2.25-inch ID breaks, in which the upper portion of the 5" core was uncovered for a longer time period relative to the rest of the core, a 19 axial node model was used to simulate the power rod. This q%g 1

permitted a detailed representation of the upper portion of the rod which experiences a relatively long heatup period. h Q

kg The limiting small break LOCA was that resulting from a safety injection 8 (2.25-inch ID) pipe rupture for which the peak clad temperature was calculated to be 18720F, occurring at 1170.2 seconds. Previous analyses d

had shown the 4.0-inch ID pipe rupture to be limiting.

M h-The reason for the shift in the limiting small LOCA break size was due $glB to the fact that a smaller break size, the 2,25-inch ID safety injection d6 oice rupture, was identified in which direct spillage of ECCS fluid could occur.

g This break results in a slower RCS depressurization such that ECCS 2 injection into the intact loops occurs much later in the accident that in @

the 4-inch break, thereby allowing a substantially longer heatuo period. 7 9

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u For the 4, 5 and 10-inch 10 breaks, depressurization was fast enough to allow the ECCS system to inject into the RCS sooner and to terminate the transient. For breaks smaller than 2.25-inches ID (size of the thermal sleeve in the ECCS penetration into the RCS), no direct spillage produced m:re effective ECCS performance.  ;

Ccoclusions a 1

From its review of the June 16, 1976, confirmatory small break analysis, ,

the Staff concludes: (1) the analysis acceptably covers the small break '

spectrum and identifies the limiting small break size, (2) the correct ECCS accumulator parameters (pressure and flow) were used in the calculations, and (3) the analysis methods and results are acceptable and do not involve a decrease in the safety margin for the Yankee-Rowe Core XII _

ECCS performance evaluation.*

Environmental Considerations J.

5 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 ;j determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental af ircact and pursuant to 10 CFR 551.5(d)(4) that an environmental impact M

statement, or negative declaration and environmental impact appraisal $

(As need not be prepared in connection with the issuance of this amendment.

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4 We have concluded, based on the considerations discussed above, that: (1)  %.

because the amendment does not involve a significant increase in the $

or:bability or consequences of accidents previously considered and does M cot involve a significant decrease in a safety margin, the amendment does not involve a significant hazards considerations, (2) there is reasonable $

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'5uosequent to the review reflected in this evaluation, we obtained new information requiring a modification of upper head water temperature M'

assumed in the ECCS evaluation for Yankee-Rowe. As a result, the NRC $$*

on August 27, 1976 issued an Order for Modification of license, restricting  %.

Erh the ceak linear heat generation rate to 0.85 kw/ft, and requiring a d' revised ECCS evaluation.

M me This amendment correc. ting accumulator pressure, does not affect the re:virement or cunclusions of the Order of August 27, 1976. The only h W

effect the Order has on considerations discussed in this safety evaluation NF would be to maintain or reduce the peak clad tempe'rature set forth herein.

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$. s assurance that the health and safety of the public will not be endanoered j by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance 7 B

of this amendment will not be inimical to the common defense and security ($

. or to the health and safety of the public. A Date: October 7,1976 s

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