ML19319C406
| ML19319C406 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 07/07/1978 |
| From: | Streeter J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Woodruff R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| NUDOCS 8002140788 | |
| Download: ML19319C406 (2) | |
Text
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a UNITED STATES na ma0g NUCLEAR REGULATORY COMMISSION
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,o July 7, 1978 Docket No. 50-346 i
MEMORANDUM FOR:
R. Woodruff, Acting Assistant Director for Technical
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Programs, Division of Reactor Operations Inspection, IE THRU:
. Fiorelli, Chief, Reactor Operations and Nuclear j
Support Branch FROM:
J. F. Streeter, Chief, Nuclear Support Section #1
SUBJECT:
INITIAL CORE SELECTIVE LOADING AT DAVIS-BESSE NUCLEAR STATION UNIT 1 (AITS F30400H2)
During an inspection which took place in the Toledo Edison (TECO) corporate offices on June 30, 1978, it was determined that a report attached to a memorandum from Babcock and Wilcox to TECO (BWT 1467, dated February 2,1977),
was not transmitted to NRR for review. The memorandum suggested that the attached report be formally submitted to the NRC and that it should reference the Densification Report (BAW-1420). A copy of the subject memorandum and report was telecopied to you on July 3,1978, for your information.
In reviewing the contents of the report we note that eight fuel assemblies l
loaded in the core do not meet the Technical Specification bases of 20.4 KW/FT for. fuel centerline melting. One of those references to the bases may be found on page B2-2 of the Technical Specifications. An additional 1
reference to the 20.4 KW/FT value may be found in section 3/4.2, Power q
Distribution Limit-Bases.
J Six of the eight assemblies have a fuel melt limit of 20.35 KW/FT and two j
have a limit of 20.17 KW/FT. Discussions with Babcock and Wilcox repre-sentatives indicated that differences in the fuel melt limits for the loaded assemblies were related to resintering the fuel pellets. As a result of the resintering process, some pellets densified differently than others and pellet to clad conductences were affected by dimensional changes.
In order to mitigate the effects of the lower fuel centerline melting limit the eight assemblies were selectively located in areas of the core which were predicted to be subjected to relatively low linear heat generation rates.
B o og 40 g
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R. Woodruff 2
July 7,1978 i
To date we have no indication that fuel centerline melting limits have been exceeded in any assembly in the core.
i Our review of the licensee's rod drop testing results are continuing. This testing indicated that extrapolation of linear heat generation rates to 100%
power conditions (with a rod drop) could result in values exceeding the
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4 20.17 limit for some assemblies, not necessarily the eight assemblies referenced. We have requested from the licensee additional information regarding the removal of conservations during the results evaluation.
j During the review of the report, we compared relative power values presented in the report with values presented in the FSAR and found that correlation was difficult. Discussions with a Babcock and Wilcox representative indi-cated that part of the explanation for the differences may be that the values presented in the selective loading report are based on calculations performed s
considering the resintering of the fuel and subsequent density and dimen-sional changes.
1, i
Since no evaluation was performed by the licensee regarding the safety is-plications of the aforementioned changes, we have requested that an eval-uation be performed prior to the escalation of power beyond the 50% power j
level.
Informal: discussions with NRR indicate minimal safety concern over the matter.
Based on discussions with the licensee and NRR, Region III does not believe this condition would effect safe operation of the facility. However, we do believe this information should have been reported to NRR as related to t
the densification issue. Region III requests that this matter be discussed with NRR to verify that this information does not affect:
a l
1.
The conclusion on the LOCA safety evaluation.
l The assurance of fuel integrity during condition I and condition II
- 2.
- accidents.
l 3.
Technical Specification Figure 2.1-2, Reactor Core Safety Limit is l
conservative.
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J. F. Streeter, Chief l
Nuclear Support Section #1 cc:
N. C. Moseley, IE J. H. Sniezek, IE IE Files Central Files T. N. Tambling, RIII J. G. Keppler, RIII x, _
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