ML17194B762
| ML17194B762 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 11/01/1983 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Farrar D COMMONWEALTH EDISON CO. |
| References | |
| LSO5-83-11-003, LSO5-83-11-3, NUDOCS 8311040083 | |
| Download: ML17194B762 (4) | |
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November 1, 1983 Docket No. 50-249 LS05-83-ll-003 DISTRIBUTION Docket File NRC PDR Local PDR NSIC OELD ELJordan JMTaylor ACRS (10)
Mr. Dennis L. Farrar Director of Nuclear Licensing Commonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690
Dear Mr. Farrar:
ORB #5 Reading DCrutchfield HSmith RGilbert
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION - USE OF ASEA - ATOM CRD BLADES IN THE DRESDEN 3 CYCLE 9 RELOAD CORE
- Re:
Dresden Nuclear Powei Station, Unit No. 3 The staff has reviewed the information, transmitted in your July 18, 1983 submittal, relating to the use of eight ASEA - Atom CRD blades in the Dresden Unit 3 Cycle 9 Reload core.
The staff finds that additional information described in the enclosure to this letter is necessary to complet~ the review.
In order to complete the review so that your schedular requirements are met, a rapid turn around is need~d. *If a meeting is required to resolve these concerns, the.information request items will serve as an* agenda.
The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.
Enclosure:
Request for Additional Information cc w/enclosure:
See next page
- Sincerely, Original signed by Dennis M. Crutchfield, Chief Operati~g Reactors Branch ~5 Division* of Licensing DL~IJJr;...
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Mr. Dennis cc Isham, Lincoln & Beale Counselors at Law One First National Plaza, 42nd Floor Chicago, Illinois 60603 Mr. Doug Scott Plant Superintendent Rural Route #1 Morris, Illinois 60450 U. S. Nuclear Regulatory Commission Resident Inspectors Office Dresden Station RR #1 Morris, Illinois 60450 Chairman Board of Supervisors of Grundy County
- Grundy County Courthouse
- Morris, Illinois 60450 U. S. Environmental Protection Agency Federal Activities Branch Region V Office ATTN:
Regional Radiation Representative 230 South Dearborn Street Chic~go, Illinois 60604 James G. Keppler, Regional Administrator*
Nuclear Regulatory Conmission, Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 Mr. Gary N. Wright, Manager Nuclear Facility Safety Illinois Department of Nuclear Safety 1035 Outer Park Drive, 5th Floor Springfield, Illinois 62704 November 1, 1983
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REQUEST FOR ADDITIONAL INFORMATION DRESDEN NUCLEAR POWER STATION, UNIT NO. 3 ASEA-ATOM CONTROL BLADES
- 1.
In the ASEA-ATOM control blades to be tested in Dresden Unit 3, the s4c absorber material will be contained in holes drilled horizontally in each blade wing.
The holes, which are closed at each end, are connected by a narrow slit that is intended to pennit pressure equalization between the holes, and the internal gas pressure and blade stress analyses are based on the assumption that pressure equalization will occur.
Are there any potential mechanisms by which blockage of the slits between the absorber-containing h~les could occur thereby resulting in excessive localized internal pressure and blade stress? If such blockage were to occur early in life, what would be the maximun internal pressure within the sealed-off hole, and what would be the corresponding stress within the blade section? Discuss the potential safety significance of this; e.g., could the blade bulge out so as to interfere with control blade insertion, or could a sudden bursting of the blade wall due to excessive internal pressure cause significant damage to an adjacent fuel assembly,
channel?
- 2.
On page 15 of ASEA-ATOM Technical Report BR 82-98, Revision 1, it is stated that a maximum internal pressure of 2030 psi can be allowed without exceeding ASME-III criteria, yet in the next sentence the selected internal design pressure is given as 2175 psi.
Why is the design pressure greater than that pennitted by the ASME code?
Specifically which ASME criteria are involved in this comparison?
Provide further detail {including report Ref. 5) regarding the methods used in the stress analysis and the means used to achieve the selected internal design pressure (see last sentence on page 15).
- 3.
The assumption of a helium release of 10 percent with no burnup dependence appears conservative with regard to internal pressure, but it could be non-conservative if used for thennal conductivity and blade temperature calculations. Please comment on the assumptions used for such calculations.
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- 4.
- 2 The text in the first paragraph on page 33 of ASEA-ATOM Technical Report BR 82-98, Revision 1 contains the statement that at 30 percent blade average burnup (said to be "typical of life limit"), the 42 percent B-10 limit of average burnup over a 3 ft. section will "generally have been reached." But on page 31, "about 42 percent" B-10 average burnup limit is said to be "implied" by a "generally used" criterion of 10 percent relative decrease of average reactivity worth over any 3 ft. segment of blade.
The use of imprecise wording such as "generally used, 11 "typical of" and "implies" in the context of a discussion of design bases and limits for He release, B-10 depletion, and burnup is confusing.
Please provid~ clear statements for the proposed design bases and limits (i.e., acceptance criteria) for B-10 depletion, blade burnup or exposure, and related factors such as He release, internal gas pressure, blade strain, etc.
- 5.
On page 34 of the ASEA-ATOM report, some control blade mechanical strength requirements are alluded to but are not actually named.
Please specify what those mechanical requirements (design bases and limits) are and show how adequate margin will be assured over the blade lifetime.
- 6.
In the.discussion of seismic perfonnance on pages 40 and 41 of ASEA-ATOM Technical Report BW 82-98, Revision 1, it is noted that during seismic testing the control rods inserted fully up to a critical value of bowing of the fuel channels and that in another series of tests with vibrating fuel assemblies, scram insertion was accomplished up to a certain critical amplitude of vibratory motion.
Provide these critical values for bowing and vibratory motion and show how these values will not be equaled or exceeded during either nonnal or accident conditions.
- 7.
With regard to the magnitude of friction forces generated between the test blade guide pads and adjacent fuel channels during a scram in-sertion, could the increased stiffness of the test blades (relative to standard blades), coupled with the use of guide pads instead of pins and 1 rollers and defonnation on channel surfaces (due to excessive contact forces), impede or prevent scram insertion under seismic or other accident conditions?