ML20214H088

Assessment of Proposed Fuel Surveillance Program for Fort St Vrain

ML20214H088
Person / Time
Site:	Fort Saint Vrain
Issue date:	05/31/1986
From:	Gruen E, Nalezny C EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC
Shared Package
ML20214H091	List: ML20214H088 ML20214P649 ML20214P673 ML20214P695
References
CON-FIN-D-6023 EGG-NTA-7247, TAC-57625, NUDOCS 8607100404
Download: ML20214H088 (26)

Text

.-.n Attachment 1 EGG.NTA-7247 4

ASSESSNENT OF THE PROPOSED FUEL SURVEILLANCE PROGRAM FOR FORT ST. VRAIN (Docket No. 50-267)

TAC No. 57625

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I INEL Lead Reviewer - E. Gruen NRC Lead Reviewer - E. Lantz INEL Programmer Mgr - C. L. Nalezny

, NRC FSV Project Mgr - K. Neitner NRC Program Mgr - M. Carrington Published May 1986 EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C, 20555 Under DOE Contract No. DE-AC07-761001575 FIN No. D6023 b .

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ABSTRACT This EG&G Idaho, Inc, report assesses the adequacy of the Fuel Surveillance Program proposed for the Fort St vrain reactor. The Fort St Vrain reactor is, operated by the Public Service Company of Colorado.

The November 1985 proposal with the changes indicated was concluded adequate.

/,; . 1 Docket (fo. 50-?67 TAC No. 57625  !

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FOREWORD This report is supplied as part of the ongoing evaluation assistance program being conducted for the U. S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation Division of Pressurized Water Reactor Licensing -8, by EG&G Idaho, Inc., NRC Technical Assistance Division.

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Docket No. 50-267 TAC No. 57625

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. CONTENTS .

ABSTRACT .............................................................. 11 FOREWORD .............................................................. 111

1. INTRODUCTION ..................................................... 1

2. REVIEW CRITERIA AND OBJECTIVES ................................... 2

3. EVALUATION ....................................................... 3

4. PROPOSED RESOLUTIONS ............................................. 10

5. CONCLUSIONS ...................................................... 12

6. REFERENCES ....................................................... 13 .

APPENDIX A--DIRECTORY OF NRC/PSC CORRESPONDENCE ....................... 16 O

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ASSESSMENT OF THE PROPOSED FUEL SURVEILLANCE PROGRAM FOR FORT ST VRAIN ,

1. INTRODUCTION .

A fuel surveillance program for the fort St. Vrain reactor has not

! been accepted for several years. An extensive program. consisting of 13 tasks was proposed by the Public Service Company (PSC) of Colorado in 1978 (Reference 1). The program was based in large part on the Department of Energy's (00E) interest in learning more about High Temperature Gas Reactor (HTGR) fuel and willingness to supply a significant amount of the

. funding for the program. As time went by the 00E reduced the amount they were willing to fund and the PSC reduced the scope of their proposed fuel surveillance programs. The progressive reduction in the proposed fuel

j. surveillance program was unacceptable to the Nuclear Regulatory Commission (NRC).

The correspondence between the PSC and the NRC was reviewed for the a years 1978 through 1985 looking for bases of agreement on which a proposal could be accepted. (A directory of this correspondence is provided in Appendix A.) The November 27, 1985 fuel surveillance proposal from the PSC was the most important (Reference 2).

1 The remainder of this report contains (a) review requirements, (b) the evaluation of correspondence in light of the review requirements, (c) conclusions concerning the adequacy of the November proposal with

] recomme'ndations,(d)referencts,and(a)theaiovementionedappendix.

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2. REVIEW CRITERIA AND OBJECTIVES ,

The adequacy of the proposed fuel surveillance program was evaluated. -

j against the criteria and objectives of the following documents: l

1. Code of Federal Reaulations,10 CFR 50, Appendix A, General Desian criteria for Nuclear Power plants -

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- Criterion 10 " Reactor Design"

- Criterion 26 ' Reactivity Control System Redundancy and Capability'

2. Updated Final Safety Analysis Report for the Fort St. Vrain Nuclear Generating Station, Vol 1 (Reference 5)

- Section 1.2.2.13 Radioactive Wastes and Radiation Protection.

- Section 3.2.2.5 Fission Product and Chemical Impurity Control

- Section 3.2.2.6 Control Rods and Drives

.- Section 3.2.3.4 Mechanical Limits for the Core and the Control. Rod Drives 3 Theproposedfuelsurveillanceprogramabstshowusthatmeasurements l

and examinations will be performed which will assure the design criteria and objectives are being met. For instance th,e fuel rod and graphite metrology must be performed to assure 'the control rods can move freely. In addition, the circulating activity must be closely monitored since a significant increase in the activity would indicate increased failure of i the fission product retention boundaries surrounding the fuel particles.

The following evaluation takes these considerations into account.. .

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3. EVALUATION The Fort St. Vrain reactor is fueled with TRISO fuel. A literat'ure search was performed for information on the manufacture, testing, and .

operation of the fuel. The fuel has'four layers consisting of porous carbon, pyrocarbon, silicon carbide, and a final layer of pyrocarbon The porous carbon absorbs and contains the gaseous and solid fission products.

The silicon carbide layer prevents the solid as well as the gaseous fission products from escaping. The outer pyrocarbon layer is the pressure boundary and holds the silicon carbide layer in compression.

In making the fuel, a very small amount (about 0.01% by count) of the particles have incomplete layers protecting the fueled kernel. Examinations of the irradiated particles show that only about 10% of the particles with incomplete layers release significant fission products. (Reference 3).

. Therefore, the contribution of a small population of defective layers to

. circulating gaseous activity is not significant in HTGR design t

(Reference 4). Additional examination of irradiated TRISO fuel is

! documented in Reference 5. Reference 5 reported that the fuel rods were in excellent condition with little or no cracking or chipping.

HTGR operations people experience a much lower radiation exposure than operators for PWRs or BWRs. The generally lower amount of radiation involved in all phases of HT6R operation as experienced by the Germans is discussed in Reference 6. The Germans have used 14 different kinds of fuel

, elements in their Advanced Gas Reactor (AVR), and state that " fuel elements l with highly enriched uranium or thorium fuel'with pyrocarbon coating showed ]

good retention czpability for all fission product,s at gas temperatures up i to 900 C. No damage of the layer occurred up to highest burn-ups." The Germans also note that the inclusion of a silicon carbide layer (like that used in FSV fuel) with the pyrocarbon layer is showing " excellent retention capability for hot gas temperature up to 950 C", but the burnup to date is too low for a final statement.

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- i Public Service Company's June, 1978, submittal of a fuel surve111ance , l program (Reference 1) included a statement that the Post Irradiation -

Examination (PIE) portion of the program was being funded by DOE and that -

any changes in funding would require changes in the described program. Th November 1985, they submitted a proposed fuel surveillance program based on their perspective of the review criteria and accounting for the decrease in DOE funding (Reference 2). Table 1 provides a comparison of both the June 1978 and the November 1985, proposed programs with an evaluation of the changes. However, in many instances PSC stated 1985 tasks differently and moved subtasks to different usin tasks. Thus the following summary evaluation is provided.

PSC wants to delete eight tasks from the 1978 fuel surveillance program for the following two reasons. First, many of the tasks being deleted are R&D and.were committed originally because DOE was willing to fund tasks on what they thought would be'the first of many HTGR's. There is no current expectation of more HTGR's being ordered and DOE is apparent 1y'no longer willing to fund these tasks. Second, the operation of ,

Fort St. Vrain has had an extremely low circulation activity history (only about 300 curies at cycle 3). This compares with the expected activity of 2630 curies based on experiments done by General Electric Co. (GEC) at the GAIL loop in the 1960's and compares with the Tech Spec Limit of 30900 curies for equilibrium cycle (Reference 7). Thus the fuel is performing extremely well and if the circulating activity stays low there is no need for extensive post irradiation examination work. Also considering the large amount of research alreAdy done-on TRISO fuel particles by others including General Atomic Corp. (GAC), GEC and Oak Ridge National Laboratory (0RNL), very little knowledge would be added by the PIE of the FSV fuel. .

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. . . . . . u-TA8tE 1. COMPARISO4 BETWEEII 1978 and 1985 FSV FUEL SINtVEILLANCE PROGtAM -

Tasks-Ig?8 Proposal Tasks-1985 Proposal Tasks Deleted Evaluation

1. Visual Inspection The fuel element will be inspected Obtain a photographic record of all sin The visual Inspection task is enhanced from i tith television cameras for signif t- vertical faces of at least 905 of the spent the 1978 proposal which is appropriate. The cant outer surface phenomena such as fuel elements removed from the core during visual Inspection is the primary mode of fuel

c;issing portions, macrocracks, large refueling using the Fuel Handling Machine surveillance.

distortions, discoloration, dust 35mm camera or the Cask Video Monitor.

deposits, or gross oxidation. The Evaluate all photographic records for The stated dependency on " equipment inspection will be done within the hot Indications of significant abnormalities . availability" for videotaping is unacceptable.

service facility at FSV and will be which could have an effect on the structural PSC should commit to perform the videotaping.

recorded on videotape. Integrity of the elements in a timely manner. Significant abnormalities are any PSC must connit to a time duration within unanticipated characteristics, the origin of which they will submit their report of which is not readily explainable, which inspection results to the NRC.

dif fer in nature from those observed in fuel element inspections conducted to date.

Elements exhibiting stains, scratches, abrasions, minor cracks and gouges typical of

. Prior segment inspection results would not be l'

considered abnormal in nature. Pronytly inform the 10tC of any significant j , abnormalities identified.which could have an effect on the structural Integrity of a fuel

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element. PSC will perform In-core visual examination of at least 12 Sepsent g H-451 fuel element surfaces during the Indicated refuelings using the Reactor Viewing Device.

The surfaces to be examined during the refuel-ing'will be those of fuel elements in Regions 3,13, and 18 that are adjacent to regions being refueled and that provide a normal (right angle la the vertical plane) viewing angle to the Reactor Viewing Device.

A report of laspection results will be prepared and submitted to the IEtc. In addition, the inspections will be videotaped ,

for record purposes, dependent upon equipment availability. ,

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i TABLE 1. (continued)

Tasks-1978 Proposal Tasks-1985 Proposal Tasks Deleted Evaluatfon

2. Graphite Block Metroloey At the time of refueling. five The wording of the PSC indicates they will not have the

, Detailed antal and diametrical .1985 proposal metrology robot device without DOE funding so l 5 length and bow measurements will be pre-characterized fuel elements will be it is not clear how they intend to perform withdrawn from the reactor and examined. Indicates much of l carried out within the hot service This emanination will include: this task is being this task. PSC should comunf t to Issue the f acility at FSV utillaing a metrology deleted particularly report to NitC within the connitted time, not robot device, which ls a remotely 4. visual examination

b. measurements to determine graphite the part requiring slaply

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operated a-y-a.e coordinate measure-

' ment system. presently under develop- dimensional changes the robot. .

Data evaluation and documentation of the PIE PSC says the 1978 PCS should commit to a brief preliminary ment et General Atomic. Data will be task is included in report to the NitC before going hack to power recorded on magnetic tape and will be results will be prepared within 12 months compared with pre-trradiation measure- after withdrawal of the elements from the the 1985 proposal. and then the detailed report issued within 12 months rather than " prepared within ments to establish irradiation induced reactor. 12 anoths."

graphits strain and bow.

3. Fuel Block Gaussa Scannine Gamma spectroscopic examination will The FTE's removed at the time of each The fuel block gauna References to dependency on DOE funding are be carried out at the reactor site refueling will be examined. This examina- scanning is deleted unacceptable. Safety is the issue, not DOE '

within the hot service f acility by Lion, which will be conducted following the due to lack of DOE funding.

systematic gamna scanning of the refueling, will include: robot funding.

a. visual inspection PCS should comunf t to a brief preliminary hirradiatedfuelblockinantaland redisl directions. A gamma scan rohot b. graphite block metrology report to the 18tC before going back to power system is presently under development c. fmeI block gauna scanning, subjett to and then the detailed report Issueci wILhIn tt Generel Atomic to perform the task the availability of DDE funding for 12 months rather than " prepared within cf fuel element positioning and robot refurbishment. 12 months."

movement in front of a collimated Ge Data evaluation and documentation of PIE (LI) detector, coupled with a results will be forwarded to the IstC as it becomes available. A report of the results

! multichannel analyzer system. Data allt be recorded on magnetic tape and will be forwarded to IstC within 12 months till be analyred for relative fission after withdrawal of the elements from the

, product distribution. This task is reactor.

. contingent on sufficient DOE funding i 13 complete the developmental efforts for the gamma scan robot.

4. Fission Gas Release From Irradiated Fuel under Storate ,

Upon receipt of an irradiated fuel Task deleted. Determination of The infor1mation this task would have produced fission gas .elease. will not be obtained from any other task.

element at General Atomic the element However, the extremely low circulating clll be put under sealed storage. Gas i

samplis will he entracted and measured activity indicates that very small fission for Kr-85 and tritium to establish product release rates would be expected.

fission gas release from irradiated in order to conf fra this. PSC should provide additional Information on fission product ,

fuel under storage, releases from stored irradiated fuel to 4

verify that these measurements are not required.

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Tasks-Ig18 Proposal Tasks-Ig85 Proposal Tasks Deleted Evaluation

5. Fisel Block Olssasses41y The fuel block will be disasses41ed at Destructive PIE's .of FTE-4 (Refueling 5) and This task is included in the 1985 proposal.

General Atomic hot cells to allow FTE-6 (Refueling 7) will be performed, ritrieval of monitor capsules and These destructive PIE's will consist of:

precharacterlied fuel rods, preferably a. fuel block disassembly from f uel stack positions containing b. non-destructive and destructive loost rods, for further destructive strain, stress and strength examina-csaminations. Fuel test arrays, tion of graphite casoonents .

containing experimental fuel from ORHL c. fluence / temperature /burnup monitor and CEA, will be extracted together esamination cith the surrounding e aphite by d. FSV reference fuel rod examination saucuttlag Lhrough the perSpheral using metallographic 1echnigues to coolant channels; these sections of confirm predicted fuel performance the test elements might be shipped to (ulth respect to kernel migration). l

.ORnL and CEA for further examination Data evaluation and decismentation of the PIE prior to shipment of the test fuel results will be forwarded to the NRC no sections, certain measurements done at later than sin (6) months from receipt of F5V sith the metrology and gamune scan the fuel elements at GA Technologies located robot systems will be repeated for in San Otego, CA. In addition. PCS cannits calibration purposes. to perform destructive PIE's on a selected w surveillsace element following any cycle during which design basis primary coolant activity limits as estabitshed by Specifica-tion LCO 4.2.8 " Primary Coolant Activity" of the Fort.3t. Vrain Technical Specifications are reac kd.

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6. Samma Spectroscopy of Individual Fuel Rods and Graphite Components fadividual fuel rods and complete Task deleted. Ganna Spectroscopy This task may be deleted because determining aulal stacks of rods as well as of Individual fuel fission product distribestion within the fuel graphite sections of the fuel element rods asal graphite element is not key to safe operation of the elli be gamma scanned to estabitsh components. reactor.

fission product distributions within the fuel element which are not ,obtain-ab12 through genna scanning of the t:tal fuel element.

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TAsLE 1. (continued) j Tasks-1978 Proposal Tasks-1985 Proposal Tasks Deleted Evaluation

7. Fuel Rod Examination and Physical ,

Property Measurements J Indtsldwal fuel rods will be inspected See Task 1. This task is now accomplished by Task I and stereoscopically for surface effects Task 2 which involve examination and cad compared with photographic records measurement, taken prior to assembly into the fuel Clements. Antal and radial dimensions t;lil be measured and compared with .

pretrradiation measurements to detefulne irradiation-induced strain. ~

. Additional physical properties like coefficient of thermal expanston,

! bonding force between fuel rods and graphite, and strength might be measured on selected rods, subject to

, availability of adequate test techniques within the GA hot cell facilities.

8. Hendestructive and Destructive

,, Strain. Stress and strength Examination of Graphite Components Selected portions of the structural The 1985 task was listed under Task 5. This task is included in the 1985 Task 5.

graphite material will be suluullted to -

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i nondestructive and destructive strain, stress and strength emaninations to .

cstab11sh residual stress distributions -

and strength margin. . Strain informa-

tion allt be used to estabitsh heat transfer gaps between graphite and fuel rods. .

9. Fluence / Temperature /surnup Monitor Eaaminatlon ,

, Selected monitor capsules will be The 1986 task was Ilsted under Task 5 This task is included in the 1985, Task 5.

disassembled and the monitors will be item c.

cnalyzed for f ast and thermal fluence, end-of-Ilfe irradiation temperatures

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. and fisslie and fertile fuel burnup and isotopic abundancies. These .

mersurements will be used to normalize -

4 rslattre fission product profiles obtained from gamma spectroscopic

, camminations to develop antal and radial power, burnup and fluence distributions within the fuel element

. for refinement of nuclear and thermal t

predict ions.

TAetE 1. (continued) i Tasks-1978 proposal Tasks-1985 proposal Tasks Deleted Evaluation

10. Ilondestructive Burnup Determination Fuel Rods

$21ected fuel rods underwent delayed Task deleted Mondestructive This task was originally Intended to identify neutron activation analysis for Burnup Determination a method other than use of. burnup monitors to trentum content prior to irradiation, on fuel rods determine burnup. The task is not necessary These rods will under go genna for safe operation of the reactor.

spectrometric examination and results util be correlated with destructive ,

burnup examinations done under item 9 to establish a nondestructive technique and to complement the ,

performance mapping of the fuel element, described In' Item 9.

11. Fuel performance Examination Fission g'as release measurements were The 1985 task Is ilsted under Task 5 TRIGA activation The metallographic examination portion of this performed on selected fuel rods prior item d. tests - task stays and is now part of Task 5 in the to essembly within General Atomic's 1985 proposal. As long as the circulating TRIGA reactor. Several of these fuel activity stays low in FSV the TRIGA fission e rods will be examined for fuel gas release measurements are not needed for performance assessment by TRIGA safe operation, fission gas release measurements and

• metallographic esaminations. TRIGA activation tests are subject to ,

availability of TRIGA reactor.  ; -

12. Simulated Core lleat e Emperience -

Selected fuel rods with preirradiation Task deleted $1mulated core This task may be deleted because the TRIGA activation tests will be heatup experiments information is available in Reference 8 submitted to core heatg simulation dated 1/24/19. In this letter PSC reports raperiments to evaluate fuel safety results of fuel particle microporosity for Ilmits. This test is subject to^ FSV fuel trradiated in the Peach Bottom development of an appropriate furnace Reactor.

I tad availability of the TRIGA reactor.

13. Destructive Fission product Esamination Graphite samples will be machined out Task deleted Destructive fission Destructive fission product emanination is not tf the graphite web between Coolant product examination necessary for safe ope, ration of the reactor, and fuel holes and measured for . The task appears to be a RI.D effort.

metallic fission product

• concentrations (e.g., ceslum and

• 3 strontium) to allow the .

characterization of fission product '.

citigation. ,

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4. PROPOSED RESOLUTION In order for the 1985 FSV Fuel ~ Surveillance Program to be acceptable" to the NRC staff, the following must be included in the program:

Task 1 - PSC must commit to a time duration within which the NRC would be informed of any significant abnormalities.

Task 1 - A fuel surveillance program cannot be dependent upon things such as video taping equipment. PSC should commit to nuking a video record of the inspections.

Task 2 - PSC must commit to a preliminary report,to the NRC before going back,to power and then issue the final detailed report within 12 months.

Task 3 - DOE funding is not relevant.and to any facet of the program. ,

PSC must commit to the task and issue the final report within 12 months of withdrawing the elements from the reactor.

l Task 4 - PSC must provide additional information on fission product

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releases from stored irradiated fuel to confirm that there is no need to perform Task 4 in the 1978 Fuel Surveillance Program.

Task 5 - This task should be divided int,o,two tasks. The last paragraph would form the first part of another task. Additional fuel surveillance will be required if the circulating activity increases to greater than twice that expected per the FSAR Table 3.7-1 (5260 curtei). If that happens then at the next shutdown, PSC must do the original 13 tasks defined in Reference 6 or propose another

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program acceptable to the NRC that would determine what caused the i

high activity. If the circulating activity exceeds the' Tech Spec '

Limit of 30900 curies then the reactor must be shut down and the original 13 tasks or another NRC approved program must be ,

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- performed to determine the cause of the high activity. In addition, PSC must identify corrective actions to prevent another similar ~

occurrence. The corrective actions must be taken before the reactor, may be restarted.

All Tasks - The references to dependency on DOE funding are unacceptable. Safety is the issue, not DOE funding.

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5. CONCLUSIONS The fueT surveillance proposal'as submitted by PSC addresses the surveillance features necessary as long as the TRISO fuel continues to exhibit low circulating activity record (See Appendix C, Attachment 1). It is unnecessary to cut up fuel blocks as long as the fuel continues to perform well. However, because of the lack of post irradiation examination for extensively irradiated fuel, the future is not assured for FSV. A clear reporting criteria is r'equired to alert the NRC when any significant change occurrs in circulating activity.

1 In addition PSC must commit to defining a special fuel inspection if the circulating activity increases dramatically relative to the expected activity and a still more intensive inspection if the activity reaches the l

Technical Specification Limit. PSC need not at this time describe the  !

special examinations that might be done. If needed the examinations would -

be developed at that time.

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6. REFERENCES .

1. Letter, J. K. Fuller (Public Service Co.) to Gammil, W. (NRC), ' Test

. Fuel Element Post Irradiation Examinatinn Program," P-78102, June 20, 1978. -

2. Letter, O. R. Lee (PSC) to H. N. Berkow (NRR), " Fort St. Vrain Fuel Surveillance Program," November 27, 1985.

3. "HTGR Fuel Technology Program Semiannual Report for the Period Ending September 30, 1982," GA-A16919 p. 3-46 (November 1982).

4. O. M. Stansfield et al., " Fuel Performance Models for High-Temperature Gas-Cooled Reactor Core Design," GA-A16982, p 111, 9 (September 1983).

5. J. W. Ketterer et al., " Capsule HRB-15A Post-Irradiation Examination Report," GA-A16758, p 111, iv (July 1984).

6. K. J. Kruger and 6. P. Ivens, " Safety-Related Experiences with the AVR Reactor," IAEA-Specialists' Meetina on Safety and Accident for Gas-Cooled Reactors. Oak Ridge, Tennessee, U.S. A 13-15 May 1985.

7. Fort St Vrain, Updated FSAR Final Safety Analysis Report, Revision 2,

. Section 3.7.

8. Letter, F. E. Swart (PSC) to T. P. Speis (NRR), " Fort St. Vrain Fuel Particle Coating Failure," P-79157, July 24, 1979.

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APPENDIX A DIRECTORY OF NRC/PSC CORRESPONDENCE G

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, DIRECTORY OF CORRESPONDENCE .

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FORT ST. VRAIN FUEL SURVEILLANCE r

Date/ SUBJECT mf address 78/06/20 PSC to NRC letter forwarding the post irradiation examination program for the proposed test fuel elements.

P-78102. -

78/06/20 PSC to NRC letter P-78103 stating PIE program ,

94033 026 was forwarded on June.20, 1978 in letter number P-78102.

78/06/23 Psc to NRC letter P-78103A letter stating gannu i 94033 020 scanning commitment was mistakenly included.

79/01/03 NRC (Speis) to PSC (Fuller) letter referring to

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. 15194'358 FSV FSAR of January 1978. Says all that is required at this time is a PSC statement that "should future funding changes require modifications to the current PIE program for standard fuel, the modifications would be j submitted to NRC for review and approval." A recommended i plan for fuel surveillance is enclosed.

l 79/01/24 PSC(Fuller)toNRC(GammillfletterP-79017 1.5194 139 requesting NRC issue a release to install fuel elements 1mmediately, i

81/11/16 PSC (Lee) to NRC (Novak) P-81254 transmitting 11220 071 and discussing PIE of fuel and reflector elements.

l 81/12/11 PSC (Lee) to NRC (Novak) P-81322 proposing j . 11372 102 reducing fuel surveillance work scope. They want to i

eliminate the destructive PIE of the segment 2 element,

. Nondestructive examination would be done for five elements as planned.

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83/01/24 NRC (Madsen) letter to PSC (Lee) responds to 17036 049 81/12/11 letter (P-81322) and calls for destructive ,'

examination.

83/02/10 PSC (Lee) letter to NRC (Collins) P-83056 17245 166 modifyingtheFuelSurveillanceProgram. Lists cmunitments contingent on DOE funding.

83/02/28 NRC (Madsen) letter to PSC (Lee) saying the 17515 355 commitment list is OK but'the contingency on DOE funding is not acceptable.

83/06/02 PSC (Brey) to NRC (Collins) P-83196 transmitting 18865 148 evaluation reports about cracked webs on the fuel elements.

84/01/03 PSC (Brey) to NRC (Collins) P-84001 transmitting -

21802 244 fuel surveillance report. The examination found the ,

performance of the fuel was acceptable.

i 84/04/06 PSC to NRC letter P-84104 with information from 24062 211 a meeting concerning cracked fuel elements.

l 84/02/15 PSC (Warembourg) to NRC (Collins) P-84053 l 22436 057 discusses Segment 3 fuel element inspection program.

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84/03/06 PSC (Warembourg) to NRC (Collins) P-84076 22660 039 discusses experience implementing the Segment 3 fuel element program including problem of under exposure on photographs.

84/04/06 PSC (Warembourg) to NRC (Collins) letter P-84104 24062 211 giving April 4 meeting notes and conclusions plus inspection program for Segment 3 fuel. -

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. 84/05/11 NRC (Johnson) to PSC (Lee) discusses conclusions ,

24507 023 of LANL concerning thermal stress cracking of fuel elements. A copy of the LANL evaluation report is ,

enclosed. Three action items are given PSC: (a) evaluate" the failure modes of a cracked fuel element under dynamic loadings and thermal stress loadings, (b) describe the fuel element inspection program to be put in'the Technical Specification, and (c) confirm a schedule for doing the above.

84/05/21 PSC (Warenbourg) to NRC (Johnson) P-84153 asks 24813 362 for more time to confirm a schedule for the above work because it must be contracted outside.

84/07/31 PSC (Warembourg) to NRC (Johnson) letter P-84246 .

26075 290 describing spent fuel inspection program.

• 84/10/19 NRC letter responding to the 84/07/31 letter' 27351 240 P-84246 by PSC. It concerns the fuel element program to be included in the Technical Specifications.

i 84/12/17 PSC letter to NRC transmitting Report 907079 - Post ,

28421 005 Irradiation Examination (PIE) of Fort St. Vrain cracked fuel elements. ,

85/01/18 PSC(Lee)toNRC(Johnson)IItterP-85005 making 28738 362 commitments to resolve NRC concerns. ,

85/01/31 PSC (GAHM) TO NE (JOHNSON) letter P-85031 29087 198 submits correction to Paragraph 5 of P-85011 Itr re Segment 3 Fuel Element Inspection Program.

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85/02/04 NRC TO PSC connenting on Report 907079. (Not in DCS) 85/03/21 PSC to NRC letter (P-85096) enclosing 40 copies 29581 001 of Report 907785 Nondestructive Examination of 62 Fuel and  ;

Reflector Elements from FSV., Claims to satisfy Segment 3 ,

and FTf-2 nondestructive examination as defined in  ;

P-84104, P-84076, and P-84053. Dates are 4/6, 3/6, and 2/15 '84 85/04/12 PSC letter to NRC with missing pages for report 907785.

(Not in DCS) 85/05/03 PSC (Lee) letter to NRC (Johnson) P-85151 30440 237 censo11 dating all previous FSP commitments. Includes modifications due to reduction in DOE funding.

85/09/18 PSC (Brey) letter to NRC (Hunter) P-85326 - States 31927 346 the commitment on Page 7 of Report 907785 was to DOE not NRC.

85/10/02 NRC (Hunter) to PSC (Lee) - Responds to PSC 32947 098 5/3/85 letter. Includes 1-1/2 pages of comments to reconcile before NRC will approve a Fuel Surveillance Prog.

85/11/27 PSC (Lee) to NRC (Berkow) letter P-85443 with 33758 073 proposed FSV fuel surveilla,nce program.

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E@ BIBUOGRAPHIC DATA SHEET EGG-NTA-7247

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3. fif Lt .s.e suet.T LS Assessment of the Proposed Fuel Surveillance Program for Fort St. Vrain . o.n ... oar cow,tino e a.omT *. 't.A May 1986

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EG&G Idaho, Inc. . ... c .=v =wan Idaho Falls, 10 83415 D6023

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.. o o.. iur.o Division of PWR Licensing - B Informal Office of Nuclear Regulatory Comission *""'******"'*"--""

Washington, D.C. 2055.5 N/A 11 suppLee.t f.mv mofte

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This EG&G Idaho Inc. report assesses the adequacy of the Fuel ,

Surveillance Program proposed for the Fort St. Vrain reactor. The Fort St. Vrain Reactor is operated by the Public Service Company of Colorado. ,

The November 1985 proposal with the changes indicated was concluded adequate.

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Attachment 2 Appendix 1 Fort St. Vrain Fuel Surveillance Program The Fuel Surveillance Program for the refuelings indicated shall consist of the following items as contained in Table 1.

1. Obtain a photographic record of all six vertical faces of at least 90% of the spent fuel elements removed from the core during refueling using the Fuel Handling Machine 35mm camera or the Cask Video Monitor. ,,

2. Evaluate in a timely manner all photographic records for indi-cations of significant abnormalities which could have an effect on the structural integrity of the elements. Significant ab-normalities are any unanticipated characteristics, the origin of which is not readily explainable, which differ in nature from those observed in fuel element inspections conducted to date.

Elements exhibiting stains, scratches, abrasions, minor cracks and gouges typical of prior segment inspection results would ,

not be considered abnormal in nature.

3. The Public Service Company of Colorado (PSC) commits to infonn the NRC within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of any significant abnormalities identified which could have an effect on the structural integrity of a fuel element.

4. At the time of refueling, five pre-characterized fuel elements will be withdrawn from the reactor and examined. This examination will include:

a) visual examination b) measurements to determine graphite dimensional changes.

Data evaluation and documentation of the post-irradiation examination (PIE) results will be provided to the NRC as it becomes available. A report of the examination results will be submitted to the NRC within 12 months after with-drawal of the elements from the reactor.

5. PSC will perform in-core visual examination of at least 12 Segment 9 H-451 graphite fuel element surfaces during the indicated refuelings using the Reactor Viewing Device. The surfaces to be examined during the refueling will be those of fuel elements in regions 3, 13, and 18 that are adjacent to regions being refueled and that provide a normal (right angle in the vertical plane) viewing angle to the Reactor Viewing Device. A report of inspection results will be prepared and submitted to the NRC.

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The fuel test elements (FTEs) removed at the time of each refueling will be examined. This examination, which will be conducted following the refueling, will include:

a) visual inspection b) graphite block metrology.

Data evaluation and documentation of the PIE results will be forwarded to the NRC as it becomes available. A report of the results will be submitted to NRC within 12 months after. with-drawal of the elements from the reactor.

7. Destructive PIES of FTE-4 (Refueling 5) and FTE-6 (Refueling 7) will be performed. These destructive PIES will consist of:

a) fuel block disassembly ..

b) non-destructive and destructive strain, stress and strength examination of graphite components c) fluence / temperature /burnup monitor examination

, d) FSV reference fuel rod examination using metallographic techniques to confirm perdicted fuel performance (with respect to kernel migration).

Data evaluation and documentation of the PIE results will be fonvarded to the NRC no later than six (6) months from receipt of the fuel elements at GA Technologies located in San Diego, CA.

If the design basis primary coolant activity limits as established by Technical Specification LC0 4.2.8. are exceeded, PSC commits to a sufficient number of PIES (comencing'with the next ref.ueling) to determine the cause of the excess actiyity levels.

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r TABLE 1 Fuel Surveillance Program Refueling No.

Item No. 4 5 6 7 8 9 10 and after 1 X X X X X X X 2 X X X X X X X 3 X X X X X X X 4 X X X X 5 X X X X X 6 X X X X 7 X X b

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