ML19220C326
| ML19220C326 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/07/1978 |
| From: | Ross D Office of Nuclear Reactor Regulation |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904300409 | |
| Download: ML19220C326 (3) | |
Text
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, og Cocket !!o. 50-320 MEF0PMDC:t FOR:
D. B. Vassallo, Assistant Director for LWRs, DF9 FRCM:
D. F. Ross, Jr., Assistant Director for Reactor Safety, DSS SUSJECT:
SAFETY EVALUATION OF PROPOSED CFRGE TO THE TECHNICAL SPECIFICATIONS FOR THREE MILE ISLMD U?l'T 2 Plant Name:
Three Mile Island, Unit 2 Occket thster:
50-320 Responsible Branct L'dR-4 and Project Manager:
H. Silver Systems Safety Branch Involved: Reacto-Systems Brarch Review Status:
Cccaleta Enclosed is the Reactor Systems CRanch review of proposed changes to the technical specifications that reflect the removal of orifice rod assemblies and the installation of restrainers on burnable poison rod assemblies at Three Mile Islani Unit 2.
We find the proposed change acceptable as discussed in the enclosure, hihinal sic.ed N D. s.nsee j'M D. F. Ross, Jr., Assistant Director for Reactor Safety Division of Systems Safety
Enclosure:
Safety Review cc:
S. Hanauer F. Schroeder R. Mattson R. Boyd D. Ross R. Baer
- 3. Yarga P. Check H. Silver K. Kniel J. McGough I. Ros:toczy T. Novak L. Phillips t'
S S. Israel W. Hodges
' ~... _ ~~ )M F. Orr G. htizetis W. Ercoks
_~ M2200am S. ?!ewberry D. Houston
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Contact:
Frank Orr, NRR g9
? I) 7 49-2759}
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NRC FOR.M 318 (9-76) NECM 0240 T u. a. novsawu rwv.,neevn a ome s ero - eme ea4
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Staff Evaluation Technical Specification Change 14 Three Mile Island, Unit 2 Cue to the recent concern over wear experienced in the holddown latch in units similar to Three :lile Island, Unit 2, causad by vibration of Burnable Poison Red Assemblies (SPRA's) and Orifice Rod Assemblies (0RA's), the applicant has proposed installation of restrainers on the BPRA's and on two modiff ad ORA's, and to remove the remaining thirty-eight ORA's.
The applichnt calculated that installation of the restrainers would reduce reactor coolant system (RCS) flow by less than 1 percent and removal of the ORA's would increase bypass ficw in the hot assembly by 1.6 percent.
Staff calculations performed during a review of similar changes for Davis Besse, Unit 1 (reference 1) confirm that these values are reasonable.
To ccmpensate for care flow distribution effects caused by the changes, the applicant has increased the primary system flow rate in the Three Mile Island, Unit 2, technical specifications (flow requirement increase of 2 percent).
The present margin in flow rate between measured and technical specification requirement ( 5 percent) would be reduced.
Because this operating margin is reduced from 5 percent to 3 percent, flow instrumentation was evaluated to assure that its accuracy is within the range of the margin.
The flow measurement system and its calibration were identified by the applicant to be identical to the system for Three Mile Island, Unit 1 and has a measurement uncertainty of about 1.5 percent.
This uncertainty is within the 3 percent margin available. We have reviewed the adequacy of the additional 2 percent RCS flow to compensate for the 1.6 percent increase in core bypass introduced by the core modifications.
It is estimated that this would provide an additicial 1.8 percent RCS flow through the core, which is greater than the 1.6 percent reduction because of the ORA bypass.
It was also reported that the limiting fuel assembly does not contain a BPRA during cycle 1 cperation.
Though this would result in higher hot assembly flow than considered in the calculations, no credit was taken for it.
The net effect of the increased flow and bjpass penalties is a slight increase in DNBR's.
89 250
DNSR-limited transients were reanalyzed considering the increased flow, trip setting adjustments, uncertainties, and rod bow penalties (for cycle 1 a CNSR criterien of 1.41 (BAW-2) accounts for rod bcw effects).
DNER values of 1.65 (BAW-2) for the 4-pump coastdown event and 1,58 (BAW-2) for a feedwater temperature decrease event were calculated. The 1-pump coast-down from 4-pump operation was identified to be the most limiting flow transient because it is used to determine the flux / flow trip set point.
Discussion with B&W indicates that the Di;SR for the 1-pump coastdown is 1.43 (BAW-2). The results for these three limiting transients are acceptable because they mest the applicable DNBR criterion,1.41 (SAW-2).
The applicant has also prorased to raise his icw RCS pressure trip setpoint frcm 1800 psig to 1900 psig. The increase in trip set point will result in larger safety margins.
We find that the technical specification change no.14 for Three Mile Island Unit 2, is acceptable because the analyses show that the modificatiens with the increased flow specifications will meet applicable criteria.
Reference:
1.
Memorandum, D. F. Ross, Jr., to D. B. Vassallo, " Safety Evaluation of Proposed Change to the Technical Specifications for Davis Besse Unit 1," June 13, 1978.
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