ML20244A655
| ML20244A655 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 04/20/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20244A651 | List: |
| References | |
| NUDOCS 7905310180 | |
| Download: ML20244A655 (22) | |
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SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AtiENDMEIIT 20 TO FACILITY OPERATING LICENSE NO. DPR-34 0F PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION DOCKET NO. 50-267 APRIL 20, 1979 d
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l-TABLE OF CONTENTS Page.
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'1.0 Introduction 1
2.0 Insertion of Test Fuel Elements Into The Core at the First Refueling 3
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3.0 Installation of PGX Graphite Surveillance l
Specimens 8
l Appendices A
Chronology of Amendments B
References 1
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1.0' INTRODUCTION ~
Fort St. Vrain, a -333 'Se high temperature gas cooled. reactor (IITGR) was ' des.igned by-the General Atomic Company and is being oper-ated by the.Public Service Company of Colorado near Platteville, Colorado.
On October 28, 1977, the Nuclear Regulatory Commission authorized ll operation of the reactor up to 70 percent of rated thermal power. All of the power ascension tests have been completed up to 70% of thermal power, which was authorized by Amendment 18. dated October 28, 1977.
This amendment deals with operations that Public Service Company of Colorado will perform'during the refueling shutdown; these operations include:
(1) insertion of eight test fuel elements into the reactor core, (2) insertion of. PGX graphite corrosion surveillance specimens into the core. reflector blocks. Technical Specification changes dealing q
with insertion of test fuel elements and PGX graphite surveillance speci-mens into the Fort St. Vrain core are identified in later chapters of l
this report along with a discussion of the corresponding safety signi-
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ficance of each insertion.
The Fort St. Vrain reactor is described in the Final Safety Analysis Report submitted for our review in November 1969. The Final Safety
. Analysis Report, as amended, formed a basis for our January 20, 1972 safety evaluation report and a first supplement, which was' issued on June 12, 1973. The operating license, DPR-34 was issued on December 21, j
1973. The operating' license has been amended twenty times, including l
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the amendment supported by this safety evaluation. A listing and brief description of the nineteen prior amendments is given in Appendix. A.
The Final Safety Analysis Report and other early documentation continues
.to support our safety reviews, as augmented by the additional information i
and the ' operational reports referenced herein.
The reactor achieved criticality on January 31,'1974, and low power physics testing was initiated. These low power tests, identified as the "A' Series"' tests, along with the "B Series," or power. ascension, tests lwere reported in accordance'with Section 7 of the Technical i
Specifications.
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Also, in accordance with the Technical Specifications, public Service of Colorado provides ". Reportable Event" reports and " Unusual Event" i
reports on safety items relating to abnormal, unusual -or unanticipated events that_ occur during the course of plant operations.
In addition to the reports received from the licensee, our safety reviews have benefitted from information on plant status and operations provided by i
the Office of Inspection and Enforcement, and by visits to the plant site p
by technical specialists to review plant records and the "as-built" con-dition of the plant. Our safety review has also included consideration i'
of comparable light water reactor experience and policies, information developed on gas cooled reactor safety under the sponsorship of the Office of Nuclear Regulatory Research, and information developed during the review 1
I of the General Atomic Standard Safety Analysis Report, GASSAR.
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2M INSERTION OF TEST-FUEL" ELEMENTSLINTO THE FORT ST. VRAIN CORE BACKGROUND _
Following preliminary discussions with the NRC staff with respect to -form and content, on January 9,1978, the Public Service Company of Colorado formally proposed to revise the Fort St. Vrain Technical Specifications to permit installation of eight test fuel elements into the reactor. This letter transmitted a pro-posed revision to Technical Specification 6.1, " Reactor Core
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- Design" and a document prepared by the General Atomic Company,
" Safety Analysis Report for Fort St. Vrain Test Elements FTE-1 through FTE-8," GLP-5494 (including Amendment 1) for NRC review and approval. On May 10, 1978 we informed Public Service Company of Colorado that the report GLP-5494 provided an acceptable basis for demonstrating the capability of the reload test elements to perform in a predictable manner, but withheld acceptance of GLP-5494 as a reference document until a description of the planned surveillance program for the fuel was provided, including a des-cription of, post irradiation examination program. We stated that a Department of Energy document, " Test Plan for FSV Fuel Test Elements," June 30, 1977, might provide an acceptable program.
On June 20, 1978 Public Sorvice of Colorado submitted a description of the post irradiation examination program as Amendment 2 to GLP-5494. On January 3,1979, following a series of discussions with the Public Service Company of Colorado, we stated that we 3
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would not require a privately funded post irradiation examination
.of the test fuel elements but that a ' firm' commitment must be made to perform-an approved surveillance program for the reference fuel, which we defined as the initial core loading.
Public Service of Colorado provided an acceptable comitment on January 24, 1979 including a commitment to a post irradiation examination program.
The scope of this program could be less than that described in
" Test Plan for FSV Fuel Test Elements," 278-FSV-5 Issue B dated June 30,:1977, but would be subject to the. review and approval of the NRC staff.
Amendment 3 to GLP-5494 was submitted on January 26, 1979 extending
' Cycle 2 core operation from 150 to 200 effective full power days.
This cycle would include the eight test fuel elements. Our review of GLP-5494 reported here does not include review of Amendment 3 or constitute approval for extension of Cycle 2 operation.
SUMMARY
OF REPORT, GLP-5494 The subject report describes eight fuel test elements (FTEs) proposed to be loaded with the first reload (" Segment 7") of the Fort St. Vrain gas cooled reactor.
It also presents the results of the analysis of the effects of the test element on plant normal operation and safety.
The fuel loading in the eight test elements is less than one percent of the loading in the core, and we have detennined that the test elements have been designed to be thermally 4
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and.neutronically compatible with 'the reference elements comprisinq
.the core load.
In ' addition to using a different type of graphite, near-isotropic H-451 graphite in place of needle-coke H-327, the test elements would contain several types of fuel particles which are not currently approved for operation in Fort St. Vrain. These include particles I
with BISO coatings, Th0 fertile kernals, UC and weak-acid-resin 2
2 (WAR) fissile kernels, and medium-enriched uranium (MEU) fuel (19.5% ' enriched U-235.instead of 93%. enrichment of the standard fissile materials). Coating thickness and kernel diameter in these test particies also differ from the standard fuel.
In some cases, a different rod curing process, called " cure-in-place,"
i will be'used in place of the reference, packed Al 0 bed process.
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.The analysis of normal operational performance addresses:
nuclear and-thernal effects; fission product release; and graphite structural
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and dimensional considerations. The report's safety analysis addresses:
the effect of specific reactivity events, such as rod withdrawals; loss of normal shutdown cooling; moisture ingress; I
permanent loss of forced cooling (LOFC) and rapid depressurization/
blowdown (both " design basis accidents"); and fuel handling accidents.
Research and development test experience, relative to FTE-1 through l
FTE-8 fuel types, is also provided as a part of the technical basis for the proposed fuel specifications and performance predictions.
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O The report concludes' that the test elements would have a very small effect on the operation of the core under all conditions.
SUMMARY
OF REGULATORY EVALUATION We have reviewed report GPL-5494 and, in addition, we have examined another report, " Test Plan for' FSV Fuel Test Elements," 278-FSV-5-Issue B, dated June 30,.197'7, for information about the planned post-irradiation examination of the test fuel. The report GLP-5494,' demonstrates analytically that the inclusion of the test elements in the Fort St. Vrain reactor should improve the. stress margins relative to the reference core and that the reduction in graphite dimensional changes should reduce column tilting and-resulting interfacial coolant flow.
Results of the thermal analysis indicate that the fuel test element inpile operating temperatures will be either essentially equal to, or less than, the reference fuel elements being replaced, with the exception of fuel test element-1, which will operate at a higher temperature than the element it replaces, but only for a 6-month period.
Fission product release calculations indicate that gaseous fission product release from the test elements would be only slightly higher than from the replaced elements even if failure rates in the test elements are higher, which is not expected. The effect on metallic fission product release is imperceptible. The report 6
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' GLP-5494 provides an acceptable analytical basis for _ demonstrating the' capability of the reload test elements to perfonn in a. pre-dictable manner under' normal conditions and postulated off-normal events.
PROPOSED TECHNICAL SPECIFICATION CHANGE We.have reviewed the proposed revision to Fort St. Vrain Technical Specification 6.1, Reactor Core Design Features. The proposed change involves the insertion of descriptive material concerning eight test fuel elements that are intended tc be installed during the first refueling, which began on February 19, 1979. We find that the descriptive material is sufficiently-detailed and accurate and that the. proposed technical. specification change is therefore acceptable.
CONCLUSIONS Based on our review of the documentation referenced in the bibliography to this report, we conclude that the eight test fuel elements described in the revision to Technical Specification 6.1 can be loaded into the Fort St. Vrain reactor without undue risk to the health and safety of the public.
We note that the loading of the remainder of the first core reloading (Segment 7) uses reference fuel and was provided for at the issuance of the Fort St. Vrain i
Operating License, DPR34 on December 21, 1973. As such, a separate application for this fuel is thereby unnecessary.
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3.0' INSERTION OF TiPE1PGP GR PHITEKTE'ST SEECIMENS INITHE FORT ST.
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VRAIN REACTOR INTRODUCTION.
On January 26, 1979 the.Public Service Company of Colorado proposed a new surveillance requirement, Technical Specification SR 5.2.22, "PGX Graphite Surveillance." This surveillance requirement provides for the insertion-and scheduled withdrawal of sixteen annular specimens of grade l PGX graphite into each of five bottom transition reflector elements.
A-safety analysis report accompanying the proposed Technical Specification describes the planned modification to the reflector element in detail and provides a safety analysis pertaining to the structural and core per-4 formance aspects.
The insertion of the PGX grade graphite test specimens into the Fort St. Vrain reactor is proposed to provide means for surveillance of the degree of oxidation corrosion that may occur when PGX is exposed to high temperature helium containing impurities of hydrogen, water vapor, carbon dioxide, and carbon monoxide at concentrations encountered during operation i
of the reactor. The oxidant quantities at temperatures equal to or greater than 1200 F are limited in the Fort St. Vrain reactor by Technical Speci-fication LCO 4.2.10, " Loop Impurity Levels, High Temperature, Limiting 1
Conditions," to oxidant concentrations totalling not more than 10 ppm.
0xidant levels are usually less than 10 ppm during power operation of the reactor but may vary depending upon plant conditions. The specimens provide i
a means of integrating the effects of time, temperature, and oxidant con-centrations in terms of total PGX corrosion.
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m BACKGROUND The purpose for' installing graphite corrosion surveillance specimens, in addition to the -present Technical Specification on the helium impurity n
concentration level, is to verify relatively recent laboratory experiments at Brookhaven National Laboratory and General Atomic Company indicating that the lifetime of the PGX core support blocks may be limited to less than' the planned lifetime of the reactor.
It should be noted that meaning-ful conclusions cannot be made at this time since the experiments performed thus far were for small specimens at laboratory conditions which may not be representative of reactor conditions. Consequently, the surveillance l
program described herein and additional research now being planned, which includes construction and operation, by General Atomic, of a high pressure test loop for larger specimens in 1980, is necessary to resolve questions related to continued PGX serviceability over the full life of the Fort St. Vrain reactor.
With respect to inmediate concerns and the present operation of the Fort St. Vrain reactor, a document was issued on April 4,1978 entitled, "NRC Staff Report on the Use of PGX Graphite in Core Support Structures Under Fort St. Vrain Operating Conditions." This report concluded that graphite oxidation does not, at present, constitute a problem affecting the health and safety of the public and that the on-going efforts of the i
licensee, General Atomic, the Department of Energy and the NRC research programs are sufficient to address the concern. This report also con-tained bibliography of correspondence, reports, and directives related to the use of PGX graphite in the core support structures under Fort St.
l Vrain operating conditions.
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l Since issuance of the April 4 report, additional meetings and dis-
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cussions have been held among the licensee, General Atomic, and the NRC staff and its consultants and contractors. Also, substantial further I
documentation has been prepared by the licensee and will be shortly forth-1 coming. Topics which are currently under review are:
(1) mechanical l
properties of oxidized PGX laboratory specimens, (2). feasibility of inserting test specimens capable of mechanical testing into the Fort St.
1 Vrain ' reactor, (3) relationships between oxidation depth and strength
'los's, (4) means of analytically monitoring oxidation mechanisms and structural consequen'ces, (5) consideration of radiation induced changes on PGX performance, (6) influence of thennal stress, and (7) requirements for experimental facilities for out-of-pile test adequately simulating reactor conditions.
In summary, the overall surveillance plan and research program for PGX graphite is not.yet fully developed. The surveillance requirement described by Technical Specification SR 5.2.22 is the first phase of a developing surveillance and research program pertaining to the long-term operation of the Fort St. Vrain reactor.
DESCRIPTION OF MODIFICATIONS AND PROGRAM Five of the transition elements of the bottom reflector will be replaced with modified elements, each containing sixteen annular specimens of PGX. graphite. The modified elements are functionally identical to the regular transition elements which are designed to merge flows from the separate 0.625 inch diameter coolant channels of each fuel column into one lo
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large channel for entry into the corresponding passage of the core support block. Both the regular transition reflector elements and the modified
- elements are made from Type H-327 graphite, the same graphite from which fuel elements are made. Each specimen is an annular cylinder 1.299 inch in outer diameter 0.625 inch in inner diameter and 0.875 inch long. A top end cap and a bottom stand-off piece of H-327 graphite position two l
specimens in tandem in each of eight channels of a reflector element, with 8,
a graphite cement used to retain the top cap in place. After the transition element assently is removed from reactor service, the graphite end cap will be broken and a special tool will be used to remove.the PGX specimens.
The original transition reflector element will be reinstalled in the reactor after renoval of the modified element assembly.
The basis for Technical Specification SR 5.2.22 identifies the specific core regions and the specimen withdrawal schedule. The proposed withdrawal i.
schedule for the modified transition elements provides for five withdrawal intervals ranging from the second refueling to the seventeenth refueling.
Should specimens at any withdrawal interval show a burnoff significantly greater than predicted, the withdrawal schedule will be adjusted to remove the modified elements at a faster rate.
EVALUATION AND DISCUSSION In the licensee's safety analysis report information is provided assessing the effect of the modification of the transition elements on their structural performance on the performance of the fuel. The primary 11 e___________
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' structural function of the transition reflector elements is to support the combined loads. from the weight of the core and the core pressure drop. -The ' enlargement of the coolant channels to a maximum diameter of 1.50 inch to accommodate the test specimens reduces the cross sectional area of the element by less than ten percent. As an example of the effect of this area reduction on strength margins, the licensee notes that the compressive strength for gross loading increases less than 33 psi. Since the FSAR value for 'the compressive strength of H-327 graphite is 4000 psi, we agree with the licensee that the reduction in cross sectional area, in order to accommodate the specimens, will result in a negligible reduction in the structural design margins of the element.
The licensee's safety analysis, report also considered cross block stresses and stresses on dowel pins. Differential thermal expansion between PGX graphite and H-327 graphite is accommodated by sizing the initial dimensions of the enlarged coolant channels and the test. specimens so that no thermally induced interference will occur.
Also, the maximum temperature variation across the element is predicted to be less than 80 F, which results in low order thermal stresses. The neutron exposure seen by the elegents is sufficiently low so that the possibility of irradiation induced dimensional changes is eliminated. The test specimens are located toward the center of the element to minimize any potential deleterious effects on dowel pin strength and the seismic capability of the modified E
elements.
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l The' licensee's report provides discussion supporting the fact that there will be no effects by the modified elements on fuel performance.
I Additional resistance to coolant flow will not be created by the test specimens so that the results of previously reviewed safety analyses and Technical Specifications will remain unaffected and there will be no requirement for revision of normal plant operational procedures or limits..The only effect is an increase in the refueling time required for t'ransition element installation, discharge, and replacement.
The' licensee considers failure of-a test specimen and subsequent blockage of a coolant channel an unlikely event but has provided an analysis of the effects should such an event occur.
By using methods from the
. FSAR and assuming complete channel blockage in all eight channe a of a modified element due to arbitrarily assumed failure of the test specimens, the fuel failed as a consequence of this blockage would increase the circu-lating gaseous radioactivity in the primary loop by about 33 percent.
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This increase would still place the circulating activity well below the FSAR design activity level. As the operating levels of the circulating activity are consistent and well known, any significant increase would be promptly indicated by the instrumentation monitr; ring the primary coolant activity.
Based on our review, we concur in this analysis.
CONCLUSIONS We find the proposed Technical Specification SR 5.2.22 acceptable and concluded that the PGX test specimens and the modified reflector transition elements, as _ described in this Technical Specification, can 13
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be~ inserted, irradiated and renoved from the Fort St. Vrain reactor without undue risk to the health and safety of the public. We base this finding on our concurrence with the licensee's analysis of. the structural, irradiation, and core performance aspects of the modified reflector elements as provided in the safety analysis report. We also. concur with the proposed withdrawal schedule taking note that the withdrawal. rate will. be increased.
if the material loss is greater than predicted.
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APPENDIX A CllRON0 LOGY OF FORT ST. VRAIN LICENSING ACTIONS
' PERTAINING lk0 PLAth! OPbh1Al[0N, SAlllY LVAl t[A1 IONS Af.1) L1Cl.NSI l
AMLNI)MLNiS Date Title January 20,.1972 Safety Evaluation by the Division of Reactor Licensing, U. S. Atomic Energy Commission in the matter of Public Service Company of Colorado - Fort St. Vrain Nuclear Generating Station, Docket No. 50-267. This document pertained to the review of the Final Safety Analysis Report prior to issuance of an operating license.
' June 12, 1973 Supplement No.1, Safety Evaluation by the Directorate of Licensing, U. S. Atomic Energy Commission in the matter of.Public Service Company of Colorado
~ Fort St. Vrain Nuclear Generating Station, Docket No. 50-267.
This document pertained to postulated high energy pipe ruptures outside containment.
December 21,'1973 License No. DPR-34 issued for the operation of the Fort St. Vrain Nuclear Generating Station.
May 17,1974 -
Safety Evaluation by the Directorate of Licensing Supporting Amendment No.1 to License No. DPR-34.
Changes the Technical Specifications by:
(1) making exceptions. to requirements for installation of secondary closures during certain initial low power physics testing, (2) revising specifications for monitoring during certain radioactive effluent releases, (3) revising specification for tendon load cell and PCRV concrete crack surveillance, (4) revising certain specifications for checks, calibrations, and testing of loop shutdown system, and (5) redefining certain administrative responsibilities and authorities of the offsite Nuclear Facility Safety Committee.
June 27, 1974 Safety Evaluation by the Directorate of Licensing Supporting Amendment No. 2 to License No. DPR-34.
Changed the Technical Specifications to revise the organization of personnel for Fort St. Vrain Nuclear Generating Station.
July 12,1974 Safety Evaluation by the Directorate of Licensing I
Supporting Amendment No. 3 to License No. DPR-34.
Changed the Technical Specifications to allow low power reactor operation with a helium environment in the reactor during Phase 1 of the power ascension program.
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7-Date Title November 11,'1974.
Safety Evaluation by the Directorate of Licensing supporting Amendment No. 4 to License No. DPR-34 Changed-the Technical' Specifications to permit revision of (1) radial power peaking factors under certain
. operating conditions and (2) the number of core regions
. allowed the maximum deviation in outlet-temperature from the average core outlet temperature.
December 19, 1974 Safety Evaluation by the Directorate of Licensing Supporting Amendment No. 5 to License No. DPR-34.
Changed the Technical' Specifications 'to permit revised staffing requirements for plant. operating shifts.
January 23, 1975 Safety Evaluation by the Division of Reactor Licensing, Supporting Amendment No. 6 to License No. DPR-34.
I Changed the Technical Specifications to permit a change
. in calibration frequency for one. adjustment of the wide range power instrumentation and added a calibration requirement for the linear range power instrumentation.
April 17,1975 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 7 to License No.
DPR-34. Changed-the Technical Specifications to permit bypass of the two-loop trouble scram-when the reactor mode switch is in the " fuel loading" position.
December 1,1975 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 8 to License No.
DPR-34.. Permitted the possession and use of additional
'l radioactive sources for the purpose' of calibration and instrument checks.
December 29, 1975 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 9 to License No.
DPR-34. Changed the Technica'i Specification to permit a reduction in the helium circulator high-speed trip when operating on water-driven Pelton turbines.
January 27, 1976 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.10 to License No.
DPR-34. Changed the Technical Specifications to permit a change in the procedures to be followed in the event of trouble with the hydraulic power system.
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.x l Date Ti tle April 15,1976 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No,11 to License No.
DPR-34. Changed the wording in the Technical Specifi-t cations to l eliminate an inconsistency in the' plant protection system 1abeling and the' Final Safety. Analysis Report.
April 26,1976 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.12 to License No.
Changed < the. Technical -Specifications to add '
surveillance' requirements for helium circulators and helium circulator Pelton wheels.
June 18, 1976 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.13 to License No.
DPR-34. Changed the Technical Specifications to:
(1) ' add requirements for operation of analytical system moisture monitors between reactor shutdown and 5 percent
. power;- also calibration frequency for these monitors-is stated; (2) revise allowable primary system impurity levels and method of specifying moisture impurity from parts per million to dew point temperature; (3) add a definition of operable dew point moisture monitors; (4) add functional checks and tests for dew point. moisture monitors; (5) revise the core reactivity status surveillance and limiting conditions for operation; (6) isolate the helium storage system from the helium circulator buffer helium system when the reactor is in operation; (7) allow bypass of plant protective system moisture monitors for testing during the startup' testing program; and (8) add reporting requirements.
June 18,1976 Safety Evaluation Supporting the Operation of the Fort St. Vrain Nuclear Generating Station Following Corrective Work for the Plant Electrical Installation and Amendment No.14 to Facility Operating License DPR-34.
June 24, 1976 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.15 to License No.
DPR-34. Changed Technical Specifications to add requirements for operability and surveillance of shock suppressors.
O q I Da te Ti tl e November 17, 1976 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.16 to License l
No. DPR 34.
Revised the section of the Technical Specifir.ations relating to administrative controls.
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December 8,1976 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.17 to License No.
i DPR-34. Temporarily revised the provisions in the Technical Specifications relating to operation of the bearing water makeup pumps in the primary coolant system.
October 28, 1977 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.18 to License No.
DPR-34 Permitted Stage 2 operation up to 70 percent of rated thennal power.
February 23, 1979 Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.19 to License No. DPR-34.
Incorporates the Fort St. Vrain Amended Security Plan.
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APPENDIX B, BIBLIOGRAPHY'FOR SECTION 2.0-
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F. E. Swart.(PSCo).to R. P.L Denise, " Request for Review of Draft Technical Specification Change - Installation of Fuel Test Elements" August 9, 1977, i
'2.
Public Service Company of Colorado,. Proposed Revision to Technical
-Specification 6.1, " Reactor Core-Design Features" (Draft), August 9, 1977.
3 General Atomic. Company, " Safety Analysis Report for Fort St. Vrain Reload-1 Test Elements FT1-1 through FTE-8, GLP-5494, June 0, 1977.
4.
R. P. Denise (NRC) to J. K. Fuller (PSCo), " Fort St. Vrain Reload
-Test Elements," November 25, 1977.
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C. K. Millen (PSCo) to R. P. Denise (NRC), " Technical Specification
. Change -_ Installation of Eight (8)' Test Fuel Elements, January 9, 1978.
This included Amendment 1 to Reference above and the final version of.the revision to Technical Specification 6.1.
6.
R. P. Denise (NRC) to J. K. Fuller (PSCo), " Fort St. Vrain Fuel Surveillance," April 6, 1978.
7.
'R. P. Denise (NRC) to J. K. Fuller (PSCo), " Fort St. Vrain Test Fuel Elements," May 10, 1978.
8.
Department of Energy, " Test Plan for FSV Fuel Test Elements, 278-FSV-5-Issue B, June 3 0, 1977.
'9.
J. K. Fuller (PSCo) to W. Gammill (NRC), " Test Fuel Element Post-Irradiation Examination Program," June 20, 1970.
10.
J. K. Fuller (PSCo) to W. Gammill (NRC), " Fort S. Vrain Fuel Surveillance," June 20 and 23,1978.
11.
T. ' P. Speis (NRC) to J. K. Fuller, " Fuel Surveillance," January 23, 1979.
12.
J. K. Fuller (PSCo) to W. P. Ganmill (NRC), " Standard fuel-Post-Irradiation Examination," January 24, 1978.
13.
J. K. Fuller (NRC) to W. P. Gammill (NRC), " Fuel Test Element Safety Analysis Report, Amendment 3," January 26, 1979.
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BIBLIOGRAPHY FOR SECTION 3'.0
- 1. ' Donald G. Schweitzer, Quarterly Report, July-September 1976, HTGR Safety Division, Brookhaven National Laboratory, Upton, New York, BNL-NUREG-50645, March 1977.
2.
P. M. Williams, " Report of Meeting on PGX Graphite, Docket No.
50-267 (Fort St. Vrain), November 10, 1977.
3.
NRC Staff Report on.the Use of PGX Graphite in Core Support Structures Under Fort St. Vrain Operating Conditions," April 4, 1978.
4.
P. M. Williams, " Report of Meeting on PGX Graphite held May'16 and 17,1978, Docket No. 50-267 (Fort St. Vrain) June 2, 1978.
5.
W. P. Gammill-to-J. K. Fuller " Graphite Surveillance Program,"
letter, June 8,1978.
6.
J. K. Fuller to W. P. Gammill (NRC), " Graphite Surveillance Program," letter and report, October 26, 1978.
7.
T. P. Speis (NRC), to J. K. Fuller, " Surveillance Program for PGX Graphite," letter and request for additional information, January 18, 1979.
- 8.. C. K.= Millen to W. P. Gammill, " Technical Specification Change SR 5.2.23," January 26, 1979.
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