ML20155A572

Forwards App a to SAR for Fort St Vrain Reload 1 Test Fuel Elements FTE-1 Through FTE-8, Consisting of post-irradiation Exam Program,In Response to 780510 Request. Approval of 780109 Submittal Requested

ML20155A572
Person / Time
Site:	Fort Saint Vrain
Issue date:	06/20/1978
From:	Justin Fuller PUBLIC SERVICE CO. OF COLORADO
To:	Gammill W Office of Nuclear Reactor Regulation
References
P-78102, TAC-57625, NUDOCS 8604090088
Download: ML20155A572 (8)

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M.F.S.C. O!377.120 TIC.*1, 3^

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-%', ', puhuc service company e ceneee P. O. Box 361, Platteville, CO 80551 June 20, 1978 Fort St. Vrain Unit No. 1 '

P-78102 Mr. William Garmill, Asst. Director Standardization and Advanecd 7,eactors Division of Project Management U. S. Nuclear Regulatory Co= mission Washington, D.C. 20555 Docket #50-267

Subject:

Test Fuel Element Post Irradiation Examination Program Gentlemen:

In your letter of May 10, 1978, it was indicated that PSC's request to install eight test fuel elements in the Fort St. Vrain reactor, submitted January 9, 1978 in correspondence P-78004, had been reviewed and had been found to be acceptable by the staff. Your letter of May 10, 1978, also indicated that formal approval to insert the test fuel elements in the reactor would be withheld pending submittal of a Post Irradiation Exam-ination (PIE) Program for these elements.

Please find attached forty (40) copies of the requested PIE Program for the test fuel elements. The ?IE program is submitted as Appendix A to the " Safety Analysis Report for Fort St. Vrain Reload 1 Test Fuel Elements FIE-1 through TTE-8" that was submitted in our correspondence dated January 9, 1978, reference nu:bar P-78004 It shculd be noted that the documented PIE Program is presently being funded by the Department of Energy (DOE). Any changes in the DOE funding would require changes in the described PIE Program.

With submittal of this requested ?IE Program for the test fuel elements all requirements for approval for insertion of these elements in the reactor hace been Jet. PSC therefore requests the NRC approve the request forwarded in our correspondence Technical SpecificationP-7500!..

chance. dated January 9, 1978, and issue the necessary If there are any questions, ;1 esse let me know. M Very truly yours, (,

I N

P

)[ .F Vice President Engineering and Planning JKFail

June 20, 1978 f Mr. William Ga.snt11 l Pasa two -

i At the-tiswt of the first, second, fet.rth, and sixth Yefuelings, '

.cne element will be returned to san Diego for deistructive PIF.,

This destrue.tive fuel PIE will consist of: *

- Selectiva gamma <scaoning for relative power, flux and burnup distributton.

- Analysis of te:tytr.ttyre, burn-up,.and flux monito:s

- Fuel rod metrology l

- Graphite :netrology -

- Puel performance iteasurements for amneha effect via metalography and TRMA activatior.

Data evaluation and docu:entatien of the PIE results will bg .

forwarded to the NRC as they become available.

As previously funded by DOE. indicated, the .sbeve described PIE program is presently being If such funding should te withdrawn or modif*.e4 before f.he

• described program is completed, the Public Service Company of Golorgdo would be available to discuss a modified post-irradiation examination p.rogram if ,

such a continuing program is necessary.

Very truly yours.

PUBLIC SERVICE CCMPAhT OF COLORM;0 J. '. Fuller Vice President Engineering and Planning JKF:il i

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GLP-5494 Anandaant 1 CONTENTS ABSTRACT . . . . . . . . . . . ................... y CLOSSARY . . . . . . . . ............ ....... ... v11

1. INTRODUCTION . . . . ...................... 1-1

2.

SUMMARY

.... ........................ 2-1

3. TEST OBJECTIVES AND DESIGN CRITERIA . ....... ...... 3-1

4. TECHNICAL DESCRIPTION OF TEST FUEL . . . ............ 4-1 4.1. Fuel Element Assembly and Location in the Core ..... 4-1 .

4.2. Fuel Particles ..................... 4-3 4.2.1. Driver Fuel .................. 4-3 4.2.2. Test Array' Fuel . .. ... .......... 4-3 4.3. Fuel Rods . .' . . . . . . . . . . . . . . . . . . . . . . 4-4 4.4. Fuel Element Graphite . . . .. ......... .... 4-5 4.5. Cure-In-Place Process . . . .. . . ..... . . . . , . . 4-6 4.6. Fluence, Burnup, and Te=perature Monitors . . . ..... 4-7

5. PERFORMANCE ANALYSIS - NOPliAL OPERATION .. . . ........ 5-1 5.1. Nuclear Analysis ........ ... ......... 5-1 5.1.1. Fuel Loadings and Burn.tble Poisons . ...... 5-1 5.1.2. Power Perturbations .. . ... ........ 5-2 5.1.3. Fluence Perturbations . ... ...... ... 5-3 5.1.4. Control Rod Worths and Reactivity Effects ... 5-3 5.1.5. Fuel Handling . . . .. ....... ..... 5-3 5.2. Thermal Analysis .................... 5-4 5.2.1. Analysis Procedure . .. . . . ......... 5-4 3 5.2.2. Analysis Results . . .. .. . ... ...... 5-6 5.3. Fission Product Release Analysis ... . ........ 5-8 I 5.3.1. Caseous Fission Product Release . . . . . . . . . 5-8 l 1

5.3.2. Metallic Fission Product Release . . . . . . . 5-10 ,  ;

5.3.3. Conclusions ... . . . .. . . . . . . . . . 5-12 -

, 5.4. Graphite Structural Analysis . . ... . . . . . . . . 5-13 l

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5.5. Graphite Dimensional Change . . . ... .... . . . . 5-16 ,

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j - GI.P-5494 Amendment 2

) TABLES (Continued)

. 4-5. H-451 graphite properties . .

4-6.

................ 4-13 Quantity of instrumentation monitors in test elements . . . . 4-14 I

5- 1.-

Fuel column and region power perturbations at BOL . . . . . . 5-18 5-2. Most severe environnent experienced by various fuel test elements ..........

................ 5-19 5-3. Nominal fuel particle designs in FSV FTE-1 through FTE-8

.. 5-20 5-4. Mean ultimate tensile strengths of unirradiated H-451

graphite ....... 1

................... 5-21 5-5. Comparison of maximum initial operating and shutdown tensile stresses in test element FTE-6 and H-327 elements of the reference FSV core ..................... 5-22 6-1.

Potential effects predictions . .....of fuel test elements on FSV FSAR accident

................... 6-15 7-1. Summary of Peach Bottom HTGR test element and surveillance work items for cores 1 and 2 .

............... 7-10 7-2. Graphites and other materials 'irradiated in capsule OG-2 .. 7-11 7-3. Summary of mechanical property data for graphites irradiated

} in capsules OG-1 and CG-2 . . . . . . . . . . . . . . . . . . 7-13 7-4. Coopleted GA irradiation tests of coated particles ..... 7-14 7-5. Number'of coated particle samples successfully tested to indicated exposure .. .

................. . 7-15 7-6.

Primary irradiation tests of UC 2 TRISO fuel . . . . . . . . . 7-16 7-7. Su= mary of WAR TRISO irradiation experience . . .... ... 7-17 7-8. j Primary irradiation tests of (Th,U)C TRISO, 2 ThC TRISO fuel 7-18 7-9. 2  !

Primary irradiation tests of Th0 BISO fuel . . .... ... 7-19 2

7-10. Primary irradiation tests of large-d'iameter Th0 7-11. 2 TRISO fuel . 7-20 Evolutionary changes in HTCR fuel red fabrication processes . 7-21 7-12. Completed GA irradiation tests of fuel rods . . ....... 7-22 A-1 Proposed PIE of Fort St. Vrain fuel test elements FTE-1 through FTE-8 ......... ..... . . ....... A-6 2 FIGURES 4-1. Standard FSV fuel ele =ent configuration . . . . ....... 4-15 4-2. Description of test arrays in FTE-2, -4, and -6 . . .... . )

4-16 4-3. FTE-5 element showing buffer fuel area .. . . ...... .

4-17 4-4 Core positions showing location of eig" test elements ...

-) 4-5.

4-18 Reference CA large HTOR fuel particles .. . . ...... . 4-19

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G1P-5494 Amendment 2 APPENDIX A PROPOSED POST-IRRADIATION EXAMINATION (PIE)  :

The proposed post-irradiation examination (PIE) work scope for FTE-1 through FTE-8 under DOE funding is presented in this appendix.

The work will be done providing adequate DOE funding and equipment is available at the time of element withdrawal from the core. The work scope is geared towards test techniques presently available or under development at General Atomic. Use of a different hot cell facility will require

'modificatior.s of the work scope. Withdrawal of DOE funding will neces-sitate modifications of part or all of the PIE program.

l l The examinations will consist of some or all of the following tasks in accordance with Table A-1.

(1) Visual Inspection The fuel element will be inspected with television cameras for significant outer surface phenomena such as missing portions =acrocracks, large distortions, discoloration, dust deposits, or gross oxidation. The inspection will be done within the hot service facility at FSV and will be recorded on videotape.

(2) Graphite Block Metrolcev Detailed axial and diametrical length and bcw measurements will be

, carried out within the hot service facility at FSV utilizing a metrology robot device, which is a remotely operated x-y-z-9 coordinate measurement system, presently under development at General Atomic. Data will be recorded on magnetic cape and will be compared with pre-irradiation

) measurements to establish irradiation induced graphite strain and bow.

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CLP-5494 Amendment 2

) (7) Jppl Rod Examination and Physical Property Measurements i

Individual fuel rods will be inspected stereoscopically for surface effects and compared with photographic records taken prior to assembly into the fuel elements. Axial and radial dimensions will be measured and compared with preirradiation measurements to determine irradiation-induced strain. Additional physical properties like coefficient of thermal expan-sion, bonding force between fuel rods and graphite, and strength might be measured on selected rods, subject to availability of adequate test tech-niques within the GA hot call facilities.

(8) Nondestructive and Destructive Strain. Stress and Strength Examination_

of Craphite Components Selected portions of the structural graphite material will be sub-mitted to nondestructive anu destructive strain, stress and strength examinations to establish residual stress distributione and stren'g th margin. Struin information will be used to establish heat transfer gaps between graphite and fuel rods.

(9) Fluence /Tennerature/3urnus Monitor Examination Selected monitor capsules will be disassembled and the monitors will be analyzed for fast and thermal fluence, end-of-life irradiation tempera-tures and fissile and fertile fuel burnup and isotopic abundancies. These measurements will be used to normalize relative fission product profiles obtained from gamma spectroscopic examinations to develop axial and radial power, burnup and fluence distributions within the fuel element for refine-ment of nuclear and ther=al predictions.

(10) Nondestructive Burnus Deter =ination on Fuel Rods Selected fuel rods underwent delayed neutron activation analysis for uranium content prior to irradiation. These rods will undergo gamma spectro-metric exa=ination and results will be correlated with destructive burnup examinations done under ite= 9 to establish a nondestructive burnup technique and to complement the perfor:ance =apping of the fuel element, described in item 9.

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GLP-5494 Amendment 2 As discussed in Chapter 4 of the FTE Safety Analysis Report, FTEs 1 ,

3, and 5 consain the same driver fuel as FTEs 2, 4, and 6, i.e., TRISO-coated Th62 and TRISO-coated UC '

fuel. 2 candidate FSV reload and lead Plant HTCR Destructive PIE of these three test elements can supply supplemental data on the performance of these fuels if a need becomes apparent from the nondestructive tion of FTEs 2, 4, andexamination

6. of these elements or from the destructive In addition, all of the test elements have M-451 type graphite blocks, which is also a candidate material for FSV reloads and for the Lead Plant HTCR.

Destructive PIE of FTEs 1, 3, 5, 7, and 8 can provide supplemental performance data for H-451 graphite if that obtained I from FTEs 2, 4, and 6 require tests at intermediate fluence levels or if -

the nondestructive examination of FTEs 1, 3, 5, 7, and 8 indicate abnormal performance.

Cure-in-place effects upon graphite performance can be established from the examination of FTE-6, FTE-7, and FTE-8, which contain either cured-in-place fuel (FTE-6) or FSV reference cured-in-bed fuel

} (FTE-7 and FTE-8).

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