ML19350A067
| ML19350A067 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 11/25/1980 |
| From: | Janecek R COMMONWEALTH EDISON CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8012010343 | |
| Download: ML19350A067 (7) | |
Text
Commonwealth Edison One First N;tional Plaza, Chicago. Ilknojs
? Aodress R; ply to: P1.st Offica 3x 767
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f Chicago, filinois 60690 November 25, 1980 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Quad Cities Station Unit 1 Cycle 6 Operation Per Provisions of 10 CFR 50.59 NRC Docket No. 50-254 Reference (a):
R. F.
Janecek letter to Director of Nuclear Reactor Regulation dated Septen.ber 2, 1980 Reference (b):
R. F.
Janecek letter to Director of Nuclear Reactor Regulation dated November 13, 1980.
Dear Sir:
Quad Cities Unit 1, which has concluded its fif th cycle of operation, is rapidly completing refueling outage activities.
This letter is to advise you of Commonwealth Edisor. Company'r, review of and plans regarding the reload core.
Table 1 provides a summary of key parameters for this reload.
The reload core was designed to perform under current nominal design parameters, Technical Specifications and related bases, and current setpoints (as modified by Reference (a)) such that:
1.
Core characteristics will be less limiting than those previously reviewed and accepted, or 2.
For those postulated incidents analyzed and reported in the Final Safety Analysis Report (FSAR) which could potentially be affected by fuel reload, reanalysis has demonstrated that the results of the postulated events are within allowable limits.
Commonwealth Edison Company has performed detailed reviews of the General Electric prepared Supplemental Reload Licensing documents and the associated i
bases and references.
Based or these reviews, safety evaluations were performed by Commonwealth Edison On-Site and Off-Site Review pursuant to the requirements of 10 CFR 50.59(a) and 10 CFR 50.59(b).
80120103'/3 g
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Director of-Nuclear Reactor Regulation November 25, 1980' Page 2 1
I As in the past, the Supplemental Reload Licensing Documents i
relieo on previously reviewed.and accepted analyses reported in the FSAR, the Generic Reload Fuel Application (NEDE-24011-P-A), and the most recent LOCA analyses for Commonwealth Edisoa's BWR-3's (NE00-24146A).
Commonwealth Edison verified that the reanalyses 4
were performed in accordance with General Electric reload analysis methodology as--described'in NEDE-24011-P-A and that the results of.
these reanalysesswere within previously reviewed and accepted limits.
. Assuming approval of revisions to DPR-29 transmitted by l
Reference (a), Commonwealth Edison has concluded that additional Technical Specification changes'are not required for initial i
operation of Quad Cities Unit 1 Cycle 6.
As discussed in Reference (b), the extended burnup cf one special test assembly will, however, require the approval of an extended MAPLHGR curve prior to operation beyond approximately 4,000 MWD /st incremental core average exposure in Cycle 6.
Commonwealth Edison On-Site and Off-Site Reviews have also concluded that no unreviewed safety questions as defined by 10 CFR 50.59 are involved with this reload.
More specifically:
- 1.
There is no. increase in the probability of occcurence or
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the consequences of an incident or malfunction of equipment important to safety previously evaluated in the safety analysis report.
i 2.
No additional accident or malfunction of a different type i
than any evaluated previously in the safety analysis report has been created, and L
3.
There has been no reduction in the margin of safety as i
defined in the basis for-any technical specification.
These conclusions are also based on a thorounh evaluation of a manufacturing nonconformance in some gadolinia-urania fuel rods in 122 of the reload assemblies which involves less than the design helium backfill pressure.
General Electric's description of the nonconformance is provided in Attachment 1.
The CECO evaluation addressed several. potential aspects of operation with low backfill pressure including:
effect on fuel centerline temperature at high local power a.
level in gadolinia rods, and b.
effect on peak cladding temperature during a LOCA' assuming operation at MAPLHGR. limits, i
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. irector of Nuclear Reactor Regulation D
November 25, 1980-Page 3 Reliability related aspects such as hydriding have also been considered.
Finally, verification of the reload core design will be performed per the standard startup physics tests normally performed at the start' of each reload cycle.. In additir.1 to Core Loading Verification, these tests will include, but not be limited to, those required by Technical Specifications such as:
l.
Shutdown Margin Demonst' ration, 2.
Control Rod Drive Scram Timing, 3.
Nuclear Instrument Calibrations, 4
Reactivity Anon.aly Surveillance, and 5.
Thermal Limits Evaluation.
1 1
The fo] lowing tests will be performed primarily for operational information as discussed'during the BWR Owners /GE meetings witn the Reactor Safety Branch in February and March of 1979:
1.
TIP Uncertainty Evaluation, i
- 2.
Critical Eigenvalue Comparison, and 3.
Control Roc Function and Subcritical Checks.
1
!j-Therefore, contingent on the NRC Staff's SER for Reference
.(a),
additional operating license amendments are not required for resumption of operation with the reload core.
Our estimated startup date is shown In' Table 1 for your information.
i One (1) signed original and thirty-nine (39) copies of this
. transmittal are provided for your use.
i
-Very truly yours, I
L F
Robert F.-Janecek Nuclear Licensing Administrator Boiling Water Reactors I
cc:
RIII Resident. Inspector - Quad' Cities 8449A
TABLE 1
'd40 CITIES UNIT 1 i
Reload 5
Cycle (N) 6' Cycle N-1 Burnup 7556 MWO/ST Projected Cycle N Bu:nup*
6684 MWD /ST Reload Batch Size 244 Reload Bundle Avg. Enrichment 2.82 Reload Bundle Design P80RB282 Estimated Startup Date 12/7/80
- to End of Full Power Capability (EOFPC) based on actual cycle N-1 Burnup 8449A
i
- GENERAL h ELECTRIC uuctria,owna SYSTEMS I? l V I S I O N GENERAL ELECTRIC COMPANY.175 CURTNER AVE.. SAN JOSE. CALIFCRNIA 95125 November 11, 1980 cc:
A. F. De Vita RLG:168-80 H. G. Lihou Commonwealth Edison Company One First National Plaza P. O. Box 767 Chicago, IL 60690 Attention:
H. E. Bliss /J. A. Silady
SUBJECT:
Quad Cities 1 Reload 5 Gadolinia Rods Gentlemen:
The purpose of this letter is to provide you with information on an occurrence at our Wilmington fuel fabrication facility regarding the pre-pressurization of gadolinia fuel rods. During routine destructive product auditing of gadolinia fuel rods in mid-October,1980, it was dis-covered that a gadolinta rod did not meet specification requirements on helium fill-gas concentration and pressure.
Follow-up investigation of this obicrvation revealed that the cause was the malfunction of the final weld 5,tation in the gadolinia fuel rod fabrication line.
The malfunction was tranifested as the premature seating of the end plug in the fuel rod prior to evacuation and back-fill resulting in the intermittent occurrence of inadequate fuel rod evacuation and helium back-fill on some gadolinia rods. Based on back-testing of archive rods from inventory, the gadolinia rod population potentially affected were those produced in the mid-June to mid-October period. Th0 fraction of archive gadolinia rods tested from-that time period which did not meet specificaf. ion requirements was approximately 18*..
Upon identification of the problem, immediate corrective action was taken to repair the malfunctioning weld station and to verify that all other weld stations were functiori ng properly. Future plans include i
provision of additional fail-safe interlock devices and improved operator aids.
An. engineering review of the potential impact of placing bundles containing some gadolinia rods with less than specificetion fill-gas in operation was performed.
It was concluded that no si]nificant impact on plant operation was expected nor would a substantial safety hazard exist. As a result, reporting of this occurrence to the NRC under 10CFR21 criteria was not considered necessary; however, the NRC will be informed on a for-information-only basis.
m GENERALh ELECTRIC H. E. Bliss /J. A. 511ady November 11, 1980 Although no significant adverse consequences are expected from opera-tion of this suspect fuel, steps have been taken to correct potentially affected bundles where feasible.
Replacement of suspect gadolinia rods in bundles already assembled and in storage at Wilmington is proceeding.
Quid Cities 1, Reload 5 was one of two reloads which contained some gadolinia rods frcm tha time period in question and which had already been shipped to the reactor sites. The suspect gadolinia rods are con-tained in 122 bundl(s of the Quad Cities I reload. As previously indicated, an engineering review concluded no significant operational impact or safety hazard was expected with the specific technical concerns described below.
4 The potential technical concerns identified for the out-of-specification fuel rods were associated with the presence of oxygen and nitrogen, and the reduced level of helium. The excess oxygen is expected to preferentially react quickly with the getter material; however, bounding calculations were performed assuming that the oxygen.' reacts entirely with the clad, entirely with the fuel, and entirely with the getter.
In all cases, the amount of oxygen available for the reaction results in only a negligible effect on the reacted materials. Therefore, the excess oxygen is not expected to result in a degradation in fuel performance.
In-reactor testing has shown that the levels of nitrogen present in the out-of-specification fuel rods will be rapidly absorbed by the fuel, cladding and getter with essentially complete absorption within a few hundred mwd /t exposure. Therefore, the 2
excess nitrogen is not expected to significantly affect fuel thermal per-formance.
The remaining concern rests with the thermal effect of the reduced helium conter.t on LOCA fuel performance.
For the Quad Cities 1, Reload 5 fuel assembly 'esign, a detailed analysis was performed employing manu-facturing int ts on the observed frequency and extent of the out-of-spect-fication condition.
The results of this analysis indicated that the increase in calculated peak cladding temperature during the accident, due-to the out-of-specification condition, is calculated to be less than 17'F with 95% confidence. A 17'F change in calculated peak clad temperature is less than the 20*F criteria identified in 10CFR50 Appendix K as the threshold of significant model changes. The 20 F criteria has also been accepted by the NRC for application in evaluation of other types of deviations affecting emergency core cooling system performance evaluations.
The current Emergency Core Cooling System performance analysis for Quad Cities I was performed assuming 156 bundles were drilled for bypass fl ow.
For the reference loading pattern incorporating Reload 5, a total of 416 bu'ndles will be drilled. The additional 260 drilled bundles are expected to more than compensate for the calculated 17'F increase in peak i
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GENERAL (p ELECTRIC H. E. Bliss /J. A. Silady November 11, 1980 cia-d temperature. Therefore, no change in the plant Technical Specifica-tion, MAPLHGR, will be required if this fuel is inserted into Quad Cities 1.
Very truly y urs,
/
R.
. Gridley,' Manager Fuel & Services Licensing rem e
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