ML20086R733

Submits Schedule & Program Description for Segment 3 Fuel Element Insps,Per 840210 Telcon.Program Will Be Performed Upon Request.Fuel Element Cracks Have No Effect on Fuel Particle Integrity

ML20086R733
Person / Time
Site:	Fort Saint Vrain
Issue date:	02/15/1984
From:	Warembourg D PUBLIC SERVICE CO. OF COLORADO
To:	Jay Collins NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
References
P-84053, TAC-57625, NUDOCS 8403010107
Download: ML20086R733 (6)

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m February 15, 1984 Fort St. Vrain Unit #1 P-04053 -

30!@[R DWKi Mr. John T. Collins )

Regional Administrator gg 2 'l @

U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 ~

d Arlington, Texas 76011 4 DOCKET NO.: 50-267

SUBJECT:

SEGMENT 3 FUEL ELEMENT INSPECTION PROGRAM

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REFERENCE:

1. P-83196, Brey to Collins, Dated June 2, 1983 9- 2. P-84001, Brey to Collins, N Dated January 3,1984

3. P-83348, Brey to Collirs, y'rj, Dated October 27, 1983 Q: 4. P-82394, Brey to Kuzmycz, "n- Dated September 15, 1982

5. P-78146, Fuller to Gammill, j

Dated September 6, 1978 c;<

Dear Mr. Collins:

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^ As a result of a telephone conversation on February 10, 1984, between the Nuclear Regulatory Commission, GA Technologies, and Public Service Company of Colorado, PSC committed to provide a schedule and program description for Segment 3 fuel element inspections. This commitment was a result cf NRC concerns on how PSC is addressing the o possibility of cracks being found in Segment 3 fuel elements. This letter serves to fulfill PSC's commitment.

HISTORY Cne-hundred-five fuel and reflector elements irradiated in core Segments 1 and_2 of the Fort St. Vrain HTGR have been visually and dimensionally examined in the Hot Service Facility (HSF) at '

Fort St. Vrain. Fifty-four elements from Segment 2 were examined in April,1982; fifty-one elements from Segment 1 were examined in July, 1979.

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Among the 105 fuel and reflector elements, two Segment 2 fuel elements (serial numbers 1-2415 and 1-0172) were unique in that they were observed to have,'or were suspected of having, cracks while being examined at FSV. As a result of this discovery, PSC elected to ship five irradiated fuel elements from Colorado to GA Technalogies' hot cell facility for detailed examination.

These elements were: 1-2415, which had been observed to have a crack at FSV; 1-0172, which was suspected of having a crack, but for which confirmation at FSV was impossible; 5-0801, which may have experienced higher than average stresses during irradiation; 1-0108, which was in' a column adjacent to 1-2415 in the same region; and 2-2693, which was the center (control- rod) column element at a core elevation corresponding to the location of 1-2415 and 1-0172. It should be noted that elements can be viewed in more detail at the GA Technologies' hot cell than in the HSF at FSV since they can be positioned for optimum viewing and: . illumination. Two Ko11morgen periscopes are also available for viewing and photographing the elements at up to 10X magnification. '

Visual examination in the GA Technologies' hot cell revealed

'that the crack in element.1-2415 extended down the entire length of the "B" face (the face adjacent to the side with the single dowel- pin which faces out to the adjacent region). The crack ran to the dowel socket at the top of the block and into the coolant hole inside the dowel socket at the bottom of the block.

Upon destructive examination of element 1-2415, the crack path was'also found to intersect a fuel hole (Reference 1).

The visual inspection of element 1-0172 revealed that a crack in

. face "B" also extended the length of the element and ran to the dowel socket at the top of the block. At the bottom of the block, the crack ran from the edge of the surface to the adjacent coolant' hole, but not to the coolant hole inside the

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dowel socket.

No' other cracks were observed in these elements or in tne other

three elements (5-0801, 1-0108, and 2-2693). For a more detailed discussion .of the visual examination results, see Reference 1.

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- 1 ANALYSIS To better characterize the observed cracks and to determine why they occurred, elements 1-2415 and 1-0172 were carefully scrutinized at_the GA Technologies' hot cell. The locations of the cracks suggested that they might have been caused by an improper fit between the dowel in element 1-0172 and the corresponding dowel socket in 1-2415. However, testing and E

measurements at GA Technologies discounted this possibility (Reference 1).

A _ metallographic - examination of a fuel rod frem element 1-2415 (from a fuel hole near the crack) was also performed. This examination showed that the fuel rod and the coated fuel particles were in good condition, and that there was no ir.dication of . fuel rod to graphite block interaction (Reference 2).

At: the present time, it is believed that the mechanism which caused the cracks was a combination of thermally induced and

-neutron fluence induced stresses. The thermal stress was due to the relatively large inter-region gap flow induced thermal gradient that existed at the edge of the elements. The neutron fluence stress was due; in part, to the fluence gradient caused by. the. control rod for the region involved (region 8) being fully inserted throughout~ fuel cycle 2 (Reference 3). Other functions may include the change in axial fuel zoning at the core mid plane and the temperature dependance of irradiation induced shrinkage. The~ probable ~cause of the cracks was presented.in Reference 3. PSC and GA Technologies also have cooperated fully with 'the Los Alamos National Laboratory's independant evaluation of the cracks on behalf of the NRC.

- As. stated in Reference 4,'PSC' continues to believe that these cracks do.not constitute a hazard to public health and safety.

The presence of the cracks did not affect the cooling geometry

'of the fuel or.the ability . of the fuel. handling machine to safely remove the fuel elements. In addition, the capability to

. insert control-material into the reactor was not affected.

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4 Maintenance of prope'r cooling geometry is the most important

safety issue which one might associate with cracking of fuel

-element webs. In this regard, the-consequences of fuel element damage resulting in coolant flow obstruction were reviewed with the .NRC in response to questions raised in 1978 concerning fluctuations at FSV. A conservat_ive, bounding assessment was made of a scenario in which coolant flow through an entire column-is blocked due to core damage. For a blockage of helium

. flow such- as this to_ occur as a -result of cracking, a catastrophic. failure of an element would have to take place. It was concluded -that the consequences of this unlikely scenario care well within those previously reviewed by NRC for other FSV accident sequences.(Reference 5).

.. SEGMENT 3 INSPECTION PROGRAM To determine if cracking of Segment 3 elements has occurred, PSC commits to the following inspection program:

~1. Photographs of all six sides of all remaining Segment 3 fuel elements (approximately 175 elements at this point in

-the refueling) will be'obtained with the FHM 35 mm camera.

The film utilized is high quality, high density special film. Processing and printing time for the film is.

approximately three (3) weeks. The film will be removed for processing as the refueling proceeds. In this respect, all pictures will be available for ' examination prior to return to power operation.

2. 'The 'above mentioned photographs will all be evaluated by knowledgeable personnel before return to power operations.

If: in the evaluatio'n process evidence of cracking is apparent, the NRC will be notified immediately. PSC will' evaluate any evidence of fuel element cracking on a case by case basis in terms'of further investigations. We will keep the NRC. abreast: of any evaluations and/or

.. investigations.

3. For two Segment 3 elements whose operational history is believed to be sinilar to the two Segment 2 elements which have previously been'.found to be cracked, an additional s visual examination of all six sides for each of the two Segment :3 elements will be performed with the FHM cask TV camera. If, for some reason, the FHM cask TV camera is not available, these elements will be examined during the

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Post-Irradiation Examination program in the HSF.

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4. A Post-Irradiation Examination (PIE) program will be conducted for 50 to 60 Segment 3 fuel and reflector

' elements in the HSF. This examination will be similar to those previously performed. The selection of elements will be based on prudent PSC and GA Technologies engineering judgment, for the purpose of detecting cracks, other signs of structural damage, or other features of interest. The selection will take into account the current understanding of the.cause of the Segment 2 fuel elements cracks.

5. PSC will make its best effort to begin the PIE program within two months of the start of power operation

(>2% power) in fuel cycle 4. In the event that this program schedule can not be met, the NRC will be informed, and -a new schedule will be provided. Based on past PIE program activities, the fuel element inspection in the HSF will be complete within 40 days after it is started.

-Again, any evidence of fuel element cracking that is

- discovered during the PIE program will be immediately reported to the NRC. A formal written report will be

. submitted as soon as it is available. '

SUMMARY

-Correspondence which we have previously submitted cn this

-subject provides assurance that.these types of cracks have had no effect on the capability of the elements to perform their design and safety function. The cracks have not affected the ability to provide core cooling, nor the ability to insert negative reactivity. The handling of the elements has not been affected. - -

At the- end of Cycle 3, the circulating activity was approximately 265 curies, as compared with the design activity of 30,000 curies. Based upon these low activity levels and upon the metallographic examination of the fuel rod from element 1-2415, it is apparent .that the cracks have had no effect on fuel particle integrity.

We believe 'that requirements that are currently in place assure the continued safe operation of' Fort St. Vrain. Specifically, the. existing Technical Specifications already govern the allowable circulating activity, as well as those actions that are required in the event that the circulating activity unexpectedly. increases. Prompt identification of fuel particle failures would be provided by routine primary coolant surveillance testing and by the primary coolant on-line monitor, RT-9301.

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.Basrd upon all'of these considerations, PSC continues to believe that the. fuel element cracks of_the types seen to date de not have any potential for adversely affecting _the health and safety of:the public.

. Although PSC believes that the conclusion drawn in the a'ove o summary

. would have been substantiated by our previously planned Segment 3

-fuel element inspection program, we commit to perform the expanded inspection program at the request of the NRC.

If you have any further questions on this subject, please contact me.

, Very_truly yours, v h 77WW Don W. Warembourg -

Manager, Nuclear Production Fort St. Vrain' Nuclear Generating Station DWW/djm

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