ML19253C622
| ML19253C622 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 11/23/1979 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| NUDOCS 7912060591 | |
| Download: ML19253C622 (7) | |
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%e UNITED STATES
['r NUCLEAR REGULATORY COMMISSION
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WASHINGTON, C. C. 20555
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November 23, 1979 Docket No. 50-312 Mr. J. J. Mattimoe Assistant General Manager and Chief Engineer Sacramento Municioal Utility District 6201 S Street P. O. Box 15830 Sacramento, California 95813
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Dear Mr. Mattimoe:
Significant wear of the Zircaloy control rod guide tubes has been observed in facilities designed by Combustion Engineering (CE).
Similar wear has also been reported in those facilities designed by Westinghouse (W).
In our letter of June 13,1978, we requested information from Babcock and Wilcox (B&W) on the susceptability of the facilities designed by B&W to guide tube wear.
The information provided by B&W by letter dated January 12, 1979 was insufficient ii for us to conclude that guide tube wear was not a significant problem in the J.
facilities designed by B&W. This was documen 7d in our letter to B&W dated August 22, 1979.
Because significant guide tube wear could impede the control rod scram i
capability, and also effect the required coolable geometry of the reactor core, we consider this wear phenomenon a potential safety concern.
Therefore, we are requesting that you provide detailed information on the wear character-istics of the control rods on the guide tubes in fuel assemblies in Rancho Seco Nuclear Generating Station.
The enclosed NRC concerns a e provided to assist you in planning your control rod and guide tube surveillance program. When you have completed your surveil-lance program plan, we request that this program be submitted for NRC review before inplementation. Although this data-gathering program may be performed on available irradiated assemblies in spent fuel pools, we find that this issue should be resolved for each facility before startup from your next scheduled refueling outage commencing after January 1,1980.
To expedite our review of your program, a meeting at NRC headquarters in Bethesda, "aryland, has been scheduled for D?cemoer 20, 1979, at 9:30 a.m.
This meeting will provide you the opportunity tc clarify and discuss the enclosed NP.C concerns and tne details of your proposed program.
Your agema c' the meeting should be arcvided by Decerber 17, 1979.
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. Should you have any questions on the type of infomation we need or scheduling requirements, please contact our Operating Reactors Branch #4 Project Manager assigned to your facility.
Sincerely, A -4 h'! '.
U Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors Enclosure cc: w/ enclosure See next page 1
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O f
Sacramento Municipal Utility District cc w/ enclosure (s):
Christopher Ellis:n, Esq.
David S. Kaplan, Secretary and Dian Grueuich, Esc.
General Counsel California Energy C:- 1ssion 6201 S Street 1111 Howe Avenue F. O. Box 15820 Sacramento, Cal n,:rnia
. 82,s Sacramento, California 95813 Ms. Eleanor Schwart:
Sacramento County California State 3ffice Board of Supervisors 500 Pennsylvania Avenue, S.,., Rm. 201 t
827 7th Street, Room 424 Washington, D.C.
20303 Sacramento, California 95814 Docketing and Service Section Office of the Secretary Business and Municipal Department U. S. Nuclear Reculatory Commission 828 I Street Washington, D.C.
20555-Sacramento City-County Library Sacramento, California 95814 Michael L. Glaser, Esq.
1150 17th Street, N.W.
Director, Technical Assessment Washington, D.C.
2C336 Division Office of Radiation Programs Dr. Richard F. Cole (AW-459)
Atomic Safety an: Licensing Board U. S. Environmental Protection Agency Panel Crystal Mall #2 U. S. Nuclear Re;ulatory Comr.ission,
Arlington, Virginic 20460 Washington, D.C.
2C555 U. S. EnvironnXntal Protection Agency Mr. Frederick J. Shon Region IX Office Atomic Safety an: Lice sing Board ATTN:
EIS COORDINATOR Panel 215 Frenont Street U. S. Nuclear Re;ulatory Commission San Francisco, r21ifornia 94111 Washington, D.C.
20555 Mr. Robert B. Borsum Timothy V. A. Dillon, Esq.
Babcock & Wilcox Suite 380 Nuclear Power Generation Division 1850 K Street, N.W.
Suite 420, 7735 Old Georgetown Road Washington, D.C.
20006
!.cthesda, Maryland 20014 James S. Reed, Esq.
Thomas Baxter, Esq.
Michael H. Remy, Esq.
Shaw, Pittman, Potts.&
Reed, Samuel & Remy t
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717 K Street, Suite 405 18 tr et, NW Sacramento, Call ornia 95814 Washi ngtc.a, DC 20036 Herbert H. Brown, Esq.
Mr. Michael R. Eaton Lawrence Coe Lanpher, Esq.
Energy Issues Ccordinator Hill, Christopher and Phillips, P. C.
Sierra G ub Legislative Office 1900 M St., NW 1107 9th St., Rc - 1320 Washington, D. C.
20036 Sacramento, CA 3551' 4
Sacramento Municipal Utility District cc w/ enclosure (s):
Atomic Safety and Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Atomic Safety and Licensing Appeal Board Panel U. S. Nucicar Regulatory Commission Washington, D.C.
20555 fir. Richard D. Castro 2231 K Street Sacramento, California 95814 Mr. Gary Hursh, Esq.
520 Capital Mall Suite 700 Sacramento, California 95814
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California Department of Health ATTN:
Chief, Environmental Radiatien Control Unit Radiological Health Section 714 P Street, Room 498 Sacramento, California 95814 e
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Enclosure NRC CONCERNS ON CONTROL R0D GUIDE TUBE WEAR IN FACILITIES DESIGNED BY B&W The B&W surveillance experience on worn control rod guide tubes, as described in their January 12, 1979 letter, consists of (a) air testing of sixteen guide tubes from an Oconee-1 15x15 fuel assembly that had experienced one cycle of operation under a control rod assembly and (b) clam-shell sectioning of two guide tubes from a 17xl7 fuel assembly that had undergone a 1000-hour flow test under a control rod assembly. As documented in our letter of August 22, 1979, we find that this experience is not sufficient to support the B&W con-clusion that there is strong evidence for the absence of wear in B&W-designed plants.
In fact, to the contrary, worn guide tubes have been observed in Crystal
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River, Unit 3 spent fuel (see BAW-1490 Rev. 1, July 1978).
Our position is further ~
based on observations made by other NSSS vendors who have found a " plant-specific" and " core-position" dependence in the observed wear. Furthermore, out-of pile flow tests have demonstrated that the wear rate is a function of several design and operating variables.
1.
Propose a post-irradiation examination (PIE) program with a schedule for its implementation and a comitment to execute the program for NRC review.
This data-gathering program should be completed expeditiously considering 4
the availability of irradiated assemblies in all B&W-designed plants.
Details of the_ surveillance plan should include the following:
a.
Methods of examination (e.g., destructive, eddy c;rrent probe, boroscope, mecharical gage) accompanied by oualification of those methods.
5.
Characterization of the examined guide tubes, including their in-core locations, EFPHs, flow rates, fluence, and wear tire under rods (control, instrument, axial-power shacing, burnable coison, startup source, and orifice).
c.
Examination of those rods (control, instru ent, axial-ower snacing, burnable Doison, startuo so:.,rce, and ori# ice) contained within the guide tubes to ider.ti#v #atigue, stress cerrcsicr 0
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. cracking, wear, denting, or any other conc'tions ti at can l
degrade their design functior, reduce tcei-design lifetime, or imoede their movement.
d.
Analysis of results including quantificati:n o' guide tube wall wear depth and distribution. This :IE or:grar. may be satisfied in part of totality by reference to data taken from another B&W designed plant (s) that uses the sa.e type of fuel assemblies.
In such case, justification r;st be given that wear in the referenced plant adequately represents that of the plant design in question.
Provide all correlations supported by your tests and discuss how these correlarions are used to predict guide tube wear during reactor operations over the fuel lifetime.
2.
Provide an evaluation on the predicted guide tube wear on the stress analyses contained in the FSAR. The evaluation snould j'
address loadings associated with Condition-1 through a etents including fuel handling accidents, control rod scears, and seismic and LOCA transients. The discussion should descr'be t".e state of stress in the worn guide tubes and how the wear a'fects the loadbearing characteristics of the worn tubes.
(Note that r.oranifor wear results in a shift of the neutral tube axis which then induces not only direct stresses but also bending stresses.) Show tha: the loadbearing capacity of the worn guide tubes satisfies the acceptance criteria for these loading events.
3.
Provide or reference all material property correlations that are used in the guide tube stress analyses. These c:rrela:icns shculd accommodate the effects of hydrogen absorption and the crocensity for hydrogen uptake in the Zircaloy guide tubes as a f nctior, of accumulative wear.
- 4.
Address the consequences of hole formation in worn guide tubes.
Consider the extent and distribution of wear to see if hole formation is possible.
If the potential for hole formation can-
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not be discounted, evaluate the impact of such noles on the guide tube integrity, control rod motior and local thermal-hydraulic performance. This evaluation should account for flow-induced vibration resulting in crack propagation and possibl: fatigue frac-ture in locally thinned areas of the tube wall. This discussion should also address the entire core residence time, both during periods of wear (under rods; i.e., control, instrument, axial-power shaping, burnable poison, startup source, and orifice) and when the tubes are not rodded.
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