ML19263C065

Proposes Tech Spec Change for Pgx Graphite Surveillance Program.Change Involves Installation of Specimens in Five Bottom Transition Reflector Elements of Core,Then Subjecting Them to Nondestructive Test

ML19263C065
Person / Time
Site:	Fort Saint Vrain
Issue date:	01/26/1979
From:	Millen C PUBLIC SERVICE CO. OF COLORADO
To:	Gammill W Office of Nuclear Reactor Regulation
References
P-79016, NUDOCS 7902020254
Download: ML19263C065 (17)

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-Janucry 26, 1979-Fort St. Vrain :

Unit No. 1 -

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Mr. William P. Gamill

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' Assist. tnt Director for Standardization

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and Advanced Reactors

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SU. S. Nuclear Regulatory Comission' tlashine, ton, ' D.C.

- 20555 Docket No. 50-26]

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SUBJECT:

Technical' Specification - Change SR 5.2.22 -

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. Dear Mr.. Garni113 EThis letter is to propose the following change to th'c Fort St.,Vrain.

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. Technical Specifications; 4

Nev:SR 5~.2.22 and Basis' '(Attachment 1)

'The Safety. Analysis Report for the 1PGX: Graphite' Surveillance Program is -

included for your inforration1(Attachment 2).

We requent your concurrence with' the proposed revisions as. coon as possibic..

If you have 'any. questions, please contact.Mr. D. W. Warembourg, (303)

"571-7436'..

Very tral ours,

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,BEIVRE:THE-3 :.

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UNITED STATES NUCLEAR' REGULATORY C01011SSIONl

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Docket N6. 50-267c 4

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PIBLIC SERVICE C012NTY OF COLORADO'

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JApplication: for Amendment to

' Appendix A of Leility~ Operating License

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License No. DPR-34

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0F THE PUBLIC SERVICE ColfANY OF COLORADO FOR THE

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' FORT ST. VRAIll NUCLEAR GENERATING STATION

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- This application _ for Amendment to Appendixf,Alof s

3 Facility Operating-License,: License No. DPR-34, -

is submitted._for NRC review and approval.'

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' PUSLIC SERVICE -C0!EANY 0F COLORADO

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~ C. K.- II111en, Senior Vice President KELLY, STANSFIELD & _O'D010! ELL -

. Bryant O'Donnel'-

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' Robert E., Thompson _

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Public Service Company Building s

.v Denver, Colorado- '80202:

Attorneys for Applicant.

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CITY A!!D COIRITY OF DENVER)'

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, fC.. KI. Hillen.. baing first duly sworn, deposes and srys; That' he io Schior Vice President,of. Operations.of Public Servien g

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Company o'f Colorado, the" Licensee herein; that he has read the '

EforeEoing Application.for Amendment to Appendix A of Facility

' Operating License an'd knows the contents - thereof, 'and that the

.atatements and matters set' forth therein are true and correct

't to the inst of his knowledge, information and belief.

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'C. K. Hillen 4

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Subscribed,and ' sworn to before ac this 3 Tb. day of ~ January,.1979..

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Witnesolmy hand and official seal.

' !!y " commission expiren t rktt 3,19 9 I

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4 ATTACHMENT 1

9 ATTACH!IENT 1

SR 5.2.22 PGX Graphite Surveillance PGX graphite surveillance specimens shall be installed into five (5) bottom transition reflector elements of the Fort St. Vrain core to provide a means for assessing the condition of the PGX graphite support blocks during operation of the reactor.

These specimens (16 per reflector element) will be installed in reflector elements as indicated in Table 1 and will be removed at subsequent refueling intervals, as indicated in Table 1, unless the progressive examination of the specimens dictate otherwise.

Upon removal, these specimens will be subjected to nondestruc-tive examination, and compared with laboratory control specimens in evaluating oxidation rates, oxidation profiles, and general dimensional charac teristics.

The results of these tests and examinations shall be utilized to assess the condition of the PGX core support blocks in the reactor and shall also be utilized to modify, as necessary, the planned removal of subsequent PGX surveillance specimens.

The results of these examinations shall be submitted to the NRC staff for review.

Basis for Specification SR 5.2.22 The PGX graphite specimens will be placed in modified coolant channels in five (5) transition reflector elements in the hottest columns of regions 22, 24, 25, 27, and 30.

The surveillance test specimens will be subjected to the primary coolant conditions as well as other reactor parameters that are normally seen by the PGX core support blocks.

Exam-ination and tests of the surveillance test specimens at regular intervals can readily be utilized to assess oxidation rates, oxidation profiles, as well as general degradation of the PGX core support blocks to adequately p redict the structural integrity of the core support blocks over the operating life of the reactor.

SR 5.2.22 PGX GRAPHITE SURs2ILLANCE Table 1 TRANSITION ELEMENT ASSEMBLY WITHDRAWAL SCHEDULE Withdrawal at Refueling Reflector Region Column Numbe r*

25 7

2 30 3

4 24 7

6 22 6

9 27 2

17

• Schedule would be adjusted to remove transition element assemblies at a faster rate s' auld specimens at any withdrawal interial show a burnoff significantly greater than predicted.

9 ATTACHMENT 2

SAFETY ANALYSIS REPORT FOR PGX GRAPHITE SURVEILLANCE PROGRAM 1.

INTRODUCTION AND

SUMMARY

This Safety Analysis Report (SAR) evaluates the installation of PGX graphite surveillance specimens into five bottom transition reflector elements of the core of the Fort St. Vrain Nuc1 car Generating Station.

It is planned to replace five currently installed transition elements with modified elements containing PGX specimens during the first reactor refueling.

These modified elements will be withdrawn af ter approximately 1, 3, 5, 8 and 16 Effective Full Power Years (EFPY) of reactor operation and the PGX specimens examined to determine total weight loss and depth of oxidation.

These experimental measurements will be used to verify or revise, as appropriate, the PGX graphite reaction rate with primary coolant chemical impurities.

This report describes the modifications planned for the transition elements, s

' the effects of installation of the PGX specimens on structural integrity and core performance, and the impact on postulated accident sequences and consequences described in the Fort St. Vrain Final Safety Analysis Report (FSAR).

2.

PLANNED MODIFICATION It is planned to replace five bottom transition reflector elements with modified transition elements which contain PGX graphite specimens.

The modified transition elements will perform all original design functions while supporting the PGX specimens so that the PGX graphite is directly exposed to the primary

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coolant. The five transition element assemblies will be placed into position during the first refueling of the reactor.

Individual transition element assem-blics will be withdrawn at regular refueling intervals according to the schedule in Table 1.

SAFETY A!!ALYSIS PIPORT FOR PGX GRAPHITE SURVEILLtdiCE PROGRAM Page 2 The five transition elements will be modified as described below, a)

Eight coolant channels in each element will be counterbored to a depth of 2.35 inches below the top of the element to accept the PGX graphite specimens as shown in Figure 1, Pg. 2 of 2, View H.

b) A standoff, fabricated from H-327 graphite, will be inserted in each counterbored coolant channel to allow retrieval of the specimens.g fromtheelementafterreactorserviceasshowninFigure2, Item [

c) Two PGX graphite specimens will be inserted in each of the eight counterbored coolant channels (16 per element) to allow exposure of the PGX graphite specimens to reactor coolant flow as shown in Figure 2, Item [.

d) A drilled graphite cap, fabricated from H-327 graphite, will be inserted in each counterbored coolant channel to retain the PGX graphite specimens and assure coolant flow through the specimen, S

W see Figure 2, Item j'.

Each graphite cap will be glued in place using F-511 graphite cement as in-dicated in Figure 2.

The P-511 graphite cement is the same cement used for glue-ing the caps on the fuel rod holes in the FSV fuel elements.

For PGX specimen identification, see Test Plan for In-Reactor Surveillance of Grade PGX Graphite, 18R-29.

After the transition element assembly is removed from reactor service, the graphite end cap will be broken at the reduced section as indicated in Figure 2 9

using a removal tool.

The tool pulls against the graphite standoff; lifting the ctandoff, the PGX specimens, and the center of the graphite cap from the transition element.

The original transition reflector element will be rein-stalled in the reactor after removal of the modified element assembly.

The PGX surveillance specimen will then be examined to determine effecta due to exposure to reactor coolant flow.

These data will be used in assessing

SAFETY ANALYSIS hEPORT FOR PGX GRAPHITE SURVEILLANCE PROGlu.M Page 3 the condition of the PGX core support blocks in the reactor and in determining the next appropriate removal date for the remaining PGX surveillance specimens.

3.

STRUCTU".AL ANALYSIS As described in the FSV FSAR, Section 3.4.1.2, the transition element serves to merge the coolant flow from the separate 0.625 inch diameter coolant channels of each column into one large channel before it enters the channel in the core support block of the core support structure.

The primary structural function of this element is to support the combined loads from the weight of the core above it and the core pressure drop.

These combined loads produce only about 30 psi compressive stress (FSV FSAR, Section 3.4.2.2) on the top surface of the transi-tion elements.

Enlarging eight coolant channels in the transition element to a maximum 1.50 inch diameter to accommodate the PGX graphite specimen will reduce the cross sectional area of the transition element by less than 10% and raise the compressive stress to less than 33 psi. As indicated in FSV FSAR, Table 3.4-1, the compressive strength of H-327 graphite is greater than 4000 psi so there is a negligibly small effect on design margin.

The PGX graph'.te specimen and the U-327 graphite transition element have different coefficients of thermal expansion. Therefore, the initial dimensions of the enlarged coolant channel and the PGX specimen are sized so that no ther-mally induced interference will occur.

The low neutron exposure seen by the transition elements eliminates the possibility of interference from irradiation induced dimensional changes.

Temperature distributions across the transition element vary approximately linearly with a maximum difference of less than 80 F across the element.

There-fore, cross block stresses due to the thermal dictribution will be lou.

Inclusion of the PGX specimen will not change the thermal distribution.

SAFETY AW1YSIS REPORT FOR PGX GRAPillTE SURVEILLANCE PROGRM4 Page 4 The PGX specimens are located toward the center of the element and are not positioned near element dowels. The specimens will not have any deleterious effects on dowel strength, therefore, the seismic capability of the transition elenents will' not be affected by addition of the PGX specimens.

The PGX specimen oxidation rate is expected to be about 0.01 inch / year sur-face burnoff.

Inspection of the PGX specimens from surveillance elements re-moved early in the program is expected to provide confirmation of this oxidation rate.

Using the expected value of 0.01 inch / year for the surface burnoff, it is estimated that the inner diameter of the specimen, originally 0.625 inches, would increase to 0.945 inches after sixteen years of residence in the core.

The remaining cylinder would then have a wall thickness of 0.167 inches.

Since there are no applied loads on the specimen, this wall thickness is more than sufficient to ensure the structural integrity of the specimen.

If the observed rate of burnoff in samples withdrawn from the core significantly exceeds 0.01 s

inch / year, the withdrawal schedule for the remaining PGX specimens would be accelerated.

4.

PERFORMANCE ANALYSIS The PGX specimens will be designed so that they will not adversely affect the performance of the fuel.

Insertion of the modified PGX transition elements has no impact on the core nuclear design. No additional resistance to coolant channel ficw will be created by installation of the specimens so the results of the thermal-hydraulic analyses presented in the FSV FSAR, Section 3.6 remain valid.

Since fuel temperatures, graphite temperatures, and coolant flow rates are unaffected, the fission product release characteristics of the fuel are unchanged and the design radionuclide inventories presented in FSV FSAR, Section 3.7 vill not be exceeded.

SAFETY ANALYSIS REPORT FOR PGX CRAPHITE SURVEILLANGE PROGRAM Page 5 Since there is no impact on core performance, installation of the PGX specimens will not require revision of plant normal operational procedures or limits.

The only effect is an increase in the refueling time required for transition element installation, discharge and replacement.

Although the failure of a PGX specimen and subsequent blockage of the cool-ant channel is considered to be an incredf bly unlikely event (see 3. above),

the effect of such an incident would be quite small.

As indicated in the FSV FSAR, Section 3.6.5.2, complete blockage of the flow in one coolant chanael would result in a maximum increase in fuel centerline temperature of 600*F.

If this increase occurred at the point of peak fuel temperature (given as 2531*F in Table 3.6-3), the resultant fuel temperature could be conservatively assumed to cause complete fuel failure in the six fuel rod stacks surrounding the affected coolant channel.

If all eight coolant channels containing PGX specimens in a modified transition element are assumed to fail and it is conservatively assumed that all of the fuel surrounding the eight coolant channels completely failed, s

'the failed fuel fraction in the core would be increased by approximately 0.1%.

Calculations have shown that the expected circulating gaseous activity in the primary loop would increase by about 33% under these conservative hypothetical circumstances.

This is still well below the FSAR design activity level.

There-fore, the design radionuclide inventories given in the FSV FSAR, Section 3.7 remain unaffected as source terms for accident analyses.

Increases in the primary coolant activity, due to PGX specimen failure, will be observed by the primary coolant fission product continuous monitor.

The present levels of primary loop circulating activity are well known and the varia-tion among readings taken at a given power level is less than 10%.

This, coupled with the relatively short response time of the primary coolant fission product

SAFETY ANALYSIS REPORT FOR PGX GRAPl!ITE SURVEILLANCE PROGRAM Page 6 monitor (approximately 2.5 minutes), ensures that warning of any significant changes in circulating e: tivity will be quickly relayed to the plant operating personnel.

5.

CONCLUSIONS The installation of the modified transition elements will not result in any increased risk to the health and safety of the public.

No unreviewed safety questfons are raised because of their installation in the reactor and there is no impact on the operability of the plant.

No changes to the limiting condi-tions for operation, safety limits, surveillance requirements, or other parts of the Technical Specifications are required.

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SAFETY ANALYS hEPORT FOR P0X GRAPl!ITE SURVEILLANCE PROGRAM Page 7 O

Tabic 1 TRANSITION, ELDIENT ASSEMBLY WITHDRAWAL SCHEDULE Withdrawal at Reflector Refueling Region Column Number 25 7

2 30 3

4 24 7

6 22 6

9 27 2

17 O

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• Schedule would be adjuated to remosc transition element assemblies at a

'fister' rate should specimens at any withdrawal interial show a burnoff signif5cantlygreaterthanpredicted.

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FIG 2 - SPECIMEN ASSEMBLY PGX GRAPHITE SURVEILLANCE PROGRAM

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