ML17325A168
| ML17325A168 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 06/15/1987 |
| From: | Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Murley T NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| AEP:NRC:0916D, AEP:NRC:916D, NUDOCS 8706230090 | |
| Download: ML17325A168 (6) | |
Text
REQULAT INFORMATION DISTRIBUTION STEM (RIDS)
ACCE'SSION NBR: 8706230090 DOC. DATE: 87/06/i5 NOTARlZED:
NO DOCKET 0 FACIL: 50-3i6 Donald C.
Cook Nuclear Poujer Planti Unit 2.
Indiana 5
050003i6 AUTHi NAME AUTHOR AFFILIATION ALEXICH> M. P.
Indiana 8c Michigan Electric Co.
RECIP. NAME REC1P IENT AFFILIATION MURLEY1 T. E.
Document Control Branch (Document Control Desk)
SUBJECT:
Application for amend to License DPR-741 adding Rept XN-NF-87-3f(P) i entitled "Steam Line Break Analysis for Unit
- 2. " Fee paid.
DISTRIBUTION CODE:
AOOID COPIES RECEIVED: LTRj ENCL CZ SIZE: 9 TITLE:
OR Submittal:
General Distribution NOTES:
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INDIANA8 NICHIGAN ELECTRIC CONPANY P.O. BOX 16631 COLUMBUS, OHIO 43216 June 15, 1987 AEP:NRC:0916D Donald C.
Cook Nuclear Plant Unit No.
2 Docket No. 50-316 License No.
DPR-74 SUBMITTAL OF STEAM LINE BREAK ANALYSIS FOR D.
C.
COOK UNIT 2 U.S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.C.
20555 Attn:
T.
E. Murley
References:
l.
- Letter, G. N. Ward (Advanced Nuclear Fuels Corp.) to U.S. Nuclear Regulatory Commission, dated May 29, 1987 (Identifier GNW:047:87).
Advanced Nuclear Fuels Corp. Report XN-NF-87-31 (P),
"Steam Line Break Analysis for D.
C.
Cook Unit 2."
3.
- Letter, M. P. Alexich (I&MECo) to H. R. Denton (U.S.
NRC), dated May 21, 1985 (Identifier AEP:NRC:0916A).
- Letter, D. L. Wigginton (U.S.
NRC) to John E. Dolan (I&MECo) dated May 21, 1986 (Transmittal letter for Unit 2 Technical Specifications Amendment 82).
5.
XN-NF-84-93(P),
"Steam Line Break Methodology for PWRs," Exxon Nuclear Co., Richland, WA, November 1984.
Dear Dr. Murley:
The purpose of this letter is to advise you that Advanced Nuclear Fuels Corporation (ANF, formerly Exxon Nuclear Co.) has transmitted to you their Report XN-NF-87-31(P), entitled "Steam Line Break Analysis for D.
C.
Cook Unit 2."
This report was submitted by ANF via their letter GNW:047:87, dated May 29, 1987.
Via this letter, we request that the report be placed on our D.
C.
Cook Unit 2 docket, No. 50-316.
In Reference 3,
we committed to having a steam line break analysis performed by the midpoint of the current Unit 2 fuel cycle, Cycle 6.
This commitment was made a condition of the NRC's safety evaluation report for 870b230090 870bi 5 PDR ADOCK 050003ib
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Dr. T.
E. Murley AEP:NRC:0916D Cycle 6 (Reference 4).
The submittal of XN-NF-87-31(P) is intended to fulfillthis commitment.
The analyses presented in XN-NF-87-31(P) demonstrate the acceptability of operation of D.
C.
Cook Unit 2 under conditions of a steam line break.
These analyses, which are based on the methodology outlined in Reference 5, simulate four transient scenarios.
The scenarios considered initiation of the transient from two operating, conditions, Hot Zero Power (HZP) and Hot Full Power (HFP).
The main differences between the two initial conditions are the presence of the delayed neutrons and the higher stored energy in the HFP case.
From both these initial conditions, the transient was then assumed to occur both with and without offsite power.
- Here, the main differences result from the coastdown of the primary coolant pumps and the effect the resulting flow coastdown has on the timing of the events occurring during the transient.
The HZP scenario with loss of offsite power was determined to be the most limiting from a minimum departure from nucleate boiling ratio (MDNBR) standpoint.
This case resulted in an MDNBR of 1.26, which is above the 1.135 MDNBR safety limit for the correlation utilized.
The HFP and HZP scenarios, with offsite power maintained for operation of the primary coolant pumps, resulted in a return to higher power levels than the scenarios where offsite power is lost.
- However, these scenarios provide substantially greater margin to the MDNBR limit because of the higher coolant flow rate.
In no scenario evaluated,
- however, was fuel failure calculated to occur as a result of penetration of the MDNBR safety limit.
The HZP scenario with offsite power available was determined to be the most limiting from the standpoint of fuel centerline melt.
This scenario results in the highest calculated linear heat generation rate, 19.7 kW/ft.
This value is within the limit of 21 kW/ft discussed in XN-NF-87-31(P).
- Thus, no fuel failures would be expected due to fuel melt.
As discussed
- above, the HFP and HZP scenarios with offsite power maintained for operation of the primary coolant pumps returned to higher power levels than the scenarios where offsite power is lost.
Although these scenarios have substantially greater margin to the MDNBR limit because of a higher coolant flow rate, the higher power levels, in combination with the highly skewed power distribution due to the assumption of a stuck rod cluster, resulted in them being limiting in regard to the fuel centerline melt criteria.
'e a'r. T.
E. Murley r AEP:NRC:0916D We are currently evaluating T/S changes that may be necessary as a
result of the XN-NF-87-31(P) analyses.
Changes that are found to be necessary will be transmitted after we receive preliminary notification that the analyses contained in ZN-NF-87-31(P) are acceptable.
Pursuant to the requirements of 10 CFR 170.12(c),
we have enclosed an application fee of $150.00 for the proposed amendments.
This document has been prepared following Corporate procedures which incorporate a reasonable set of controls to insure its accuracy and completeness prior to signature by the undersigned.
Very truly yours, M.
. Alex ch Vice President Enclosure cc:
John E. Dolan W.
G. Smith, Jr.
- Bridgman R.
C. Callen G. Bruchmann G. Charnoff NRC Resident Inspector
- Bridgman A. B. Davis
- Region III
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