ML19322A267
| ML19322A267 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/16/1979 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Herbein J METROPOLITAN EDISON CO. |
| References | |
| NUDOCS 7903140665 | |
| Download: ML19322A267 (4) | |
Text
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44 NUCLEAR REGULATORY COMMISSION s 'i )
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WASHINGTON, D. C. 20555 of February 16, 1979
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- . s Docket No.:
50-289 Mr. J. G. Herbein Vice President Metropolitan Edison Company P. O. Box 542 Reading, Pennsylvania 19640
Dear Mr. Herbein:
By letter dated December 28, 1978, you requested amendment of Appendix A to Facility Operating License DPR-50 for Three Mile Island Unit No.1 to permit operation following refueling for Cycle 5.
Based on our review of your request to date we find we need additional information in order to complete our review.
The specific information needed is listed in the enclosure and was transmitted informally to Mr. Ron Stevens of your staff on February 9,1979, In order that we may continue our review on a schedule consistent with your projected restart date, you are requested to submit the requested information by February 23, 1979.
Sincerely, Q4AhtmJ Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors
Enclosure:
Request for Additional Information cc w/ enclosure:
See next page k
o 7s03140L(f
o Metropolitan Edison Company cc:
G. F. Trowbridge, Esquire Shaw, Pi ttman, Potts & Trowbridge 1800 !! Street, N.W.
Washington, D.C.
20036 GPU Service Corporation Richard W. Heward, Project Manager Mr. E. G. Wallace, Licensing Manager 260 Cherry Hill Road Parsippany, New Jersey 07054 Pennsylvania Electric Company Mr. R. W. Conrad Vice President, Generation 1001 Broad Street Johnstown, Pennsylvania 15907 fliss liary V. Southard, Chairman Citizens for a Safe Environment P. O. Box 405 Harrisburg, Pennsylvan'ia 17108 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 420, 7735 Old Georgetown Roao Bethesda,11a ryland 20014 Government Publications Section State Library of Pennsylvania Box 1601 (Education Building)
Harrisburg, Pennsylvania 17126 e
9
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TMI-l COMMENTS TO CYCLE 5 RELOAD REPORT (GQL 2068, December 28,1978) 1.
The control rod group withdrawal anticipated operational occurrence analysis shown in the FSAR is based (nominal conditions) on assumed values of the moderator and doppler temperature coefficients less adverse than values of these coefficients predicted to occur during the forthcoming cycl e.
Sensitivity studies shown in the FSAR show the effect of the more negative doppler coefficient but do not span the range of the anticipated moderator temperature coefficient.
Please explain how the system trip setpoints afford plant protection (DNBR and RCS pressure) at these more adverse conditions.
Please consider the full ra.nge of bank worths in your analysis.
2.
The dropped rod anticipated operational occurrence analysis shown in the FSAR is based on a doppler coefficient less adverse than the value predicted for the forthcoming Cycle 5.
Explain why you consider the FSAR analysis bounding for Cycle 5.
Provide the post rod drop peak enthalpy rise assumed in the FSAR analysis and predicted for Cycle 5.
l 3.
The ejected rod accident analysis presented in the FSAR is based on values of seff larger (less adverse) than predicted for Cycle 5.
Explain why you consider the FSAR analysis is bounding for Cycle 5.
Furthermore, confirm that the post ejected value of the peak linear heat rate assumed in the FSAR analysis bounds the Cycle 5 predicted values.
4.
Please confirm the applicability of BAW-1461, " Reactivity Insertion Assumptions Used in Safety Analysis Calculations" to the analysis of TMI-1, Cycle 5.
If applicable, show that sufficient margin will exist during Cycle 5 to accommodate the 0.09 DNBR reduction during the hypothetic four-pump q
coastdown sited in BAW-1461.
5.
Please confirm that clad collapse calculations for Cycle 5 lI were perfonned using the CROV coniputer code and associated standard modeling techniques.
6 6.
Provide the analytic bases for the revision of Technical Sper.ification 3.2.2 which would increase the minimum boric acid mix tank level from 800 ft3 to 906 ft,3 l
l
e 7.
Cycle 5 values of permitted axial power shaperod, APSR's, position vs. core power, shown as proposed Figure 3.5-2H of the plant Technical Specifications, will require long term insertion of the APSR during rated power production.
The APSR's is to be inserted no less than 6.1% into the 1
core, nor no more than 45% into the core, during operation of greater than 92% of rated power.
Please provide predicted values of fan and Fq following long term operation with the APSR's inserted to the maximum insertion limit and sub-sequent withdrawal of the APSR's to the minimum insertion limit.
8.
Figure 5-1 of your Cycle 5 reload report shows the beginning of cycle predicted planar power distribution whith APSR's inserted.
Does this calculation (a two dimensional PDQ07 calculation) represent the APSR's as if they were full length, full strength rods, or have cross sections been adjusted to represent the reduced length of the APSR's?
9.
Please confirm the applicability of BAW-10121P, "RPS Limits and Setpoints," to TMI-1, Cycle 5.
10.
Please provide the quantitative, rather than qualitative, bases for your revision of the bypass flow to 10.4% of total flow to accommodate the affect of orifice rod assembly removal.
11.
Please provide the quantitative, rather than qualitative, bases of your review of the peak enthalpy rise, Fag, from 1.78 to 1.71 to accommodate the revised bypass flow.
Response to items 10 and 11 may be made by ' specific reference to approved methods and models.
- 12. Are Figures 8-1 and 8-2, Core Protection Safety Limits, Trip Setpoint for Nuclear Overpower Based on RCS Flow and Axial Power Imbalance, respectively based on an assumed F H Of 4
1,71 or 1.78?
- 13. Table 1 of your submittal shows that safety limits calculated for Cycle 5 are less restrictive than the proposed Technical Specification Safety Limits (SL).
By inference you assert that the Limiting Safety System Setpoints (LSSS) corresponding to Technical Specification SL are more restrictive than the LSSS that would correspond to the Cycle 5 SL.
Please confirm this assertion.
Consider transient DNBR degradation during the course of postulated transients for which the LSSS are to provide protection, as well as steady state conditions used I
to determine the SL.
14.
Please commit to provide a startup test report.
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