ML20042F027
| ML20042F027 | |
| Person / Time | |
|---|---|
| Site: | Armed Forces Radiobiology Research Institute |
| Issue date: | 04/30/1990 |
| From: | Irving G ARMED FORCES RADIOBIOLOGICAL RESEARCH INSTITUTE |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20042F028 | List: |
| References | |
| NUDOCS 9005070123 | |
| Download: ML20042F027 (10) | |
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10C FA 50. 90 i
. DocpEr 30-170 l DEFENSE NUCLEAR AGENCY ,
ARMED FORCES RADioDioLOGY RESE ARCH INSTITUTE BETHESD A. M ARYLAND r0814 5145 T- '
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SUBJECT:
Requent for Changes to AFRRI Facility License April 30,1990 No. R-84, Paragraph 2.C.(1) and the Technical Specifications for License No.
R 84.
i Document Control Desk i .
U.S. Noclear Regulatory Commission l
Wuhington, D.C. 20555 l
Gentlemen: j The Armed Forces Radlebiology Research Institute (AFRRI) TRIGA reactor facility and the Technical Specifications of license R- .
requests 84 to reflecttochanges amendnecessary its licenseto (R al 84) low for enhanced operations of the AFRRI l
TRIGA reactor facility. These changes involve the following:
- An increase in the maximum licensed steady state power level from 1000 l kilowatts to 1100 kilowatts. This change is requested to make the maximum l
licensed power level specified in section 2.C.(1) of the AFRRI TRIGA l facility license commensurate with the maximum t.llowed power level of 1.1 megawatts as stipulated in Technical Specification 3.1.1 of license R-84.
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- The modification of parts of the Technical Specifications of license R-84 to allow for the implementation of fuel follower control rods.
- The addition of Console System Computer (CSC) and Data Acquisition Computer (DAC) watchdog ecrams to the Technical Specifications of license R-84 to allow for the future inste.llation of the General Atomics microprocessor based instrumentation and control system for the AFRRI TRIGA reactor.
- The clarification of the transfer of responsibilities of the Reactor Facility Director (RFD) during periods of his/her absence. This will involve the modification of Technical Specification 6.1.2 and also serves as AFRRI's proposed final action to unresolved item 170/88-04 05 from NRC inspection l
50-170/88-04.
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- Minor changes to the Technical Specifications of license R-84 to correct grammatical, style, spelling, or punctuation errors or to reflect changes in operational terminology. .
i 9005070123 900430 +
I PDR ADOCX 05000170 l P PDC ,
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. The full. detail of the proposed changes to. the Technical Specification and the associated safety analyses are provided in attachment 1. Proposed Changes to '
Technical Specifications, Armed Forces Radiobiology Research Institute. The safety i issues concerning the installation and operation of the fuel follower control rods are discussed in detail in attachment 2, Safety Analysis of FFCR Implementation and i Operation in the AFRRI TRIGA Reactor. In addition, proposed revised pages for ,
the Technical Specincations for license R-84 (attachment 3) are provided for your convenience. ;
i All proposed changes have been reviewed and approved by the AFRRI Reactor and Radiation Facility Safety Committee. If you have any questions or comments, please contact the Reactor Facility Director, hir. Af ark hicore, or the Reactor-Exceutive Officer,1st Lt hiatt Forsbacka, at (301) 2951290. Thank you for your .
Cooperation.
t Sincerely, i J g ,
m '1 :
i GEOR -
V. IRVING, I l Colone , SC, USAF Direct > ;
Enclosures:
as stated l cc: USNRC - Region I - Project Engineer Division of Reactor Projects USNRC - Headquarters Project hinnager >
Nuclear Reactor Regulation -
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i ATTACHhiENT 1 :
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PROPOSED CHANGES TO TECHNICAL SPECIFICATIONS !
ARhfED FOftCES R ADIOBIOLOGY RESEARCH INSTITUTE :
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t Armed Forces Radiobiology Research Institute :
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- hiarch 1990 F
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. 1. Pace i. Tehle of Contents. Section 1.25: Change from " PAD" to " ROD" to correct typographical error.
Safety Analysis: This is an administrative change to correct spelling.
There are no safety implications.
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- 2. Parc 1. Table of Contents: Adjust numbering of definitions to reflect addition of definition of Reactor Facility Director (1.17).
Safety Analysis: There are no safety implications.
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- 3. Pare 2. Se tion 1.9: Replaec the definition for a fuel element in its entirety to accommodate fuel follower control rods as follows: ,
1.9 FUEL ELEMENT ,
A fuel element is a single TRIGA fuel rod, or the fuel portion of a fuel ,
follower control rod. '
Safety Analysis: See attached saf(ty analysis on FFCR. ,
- 4. Pace 3. Section 1.17: Add this section in its entirety with the following:
1.17 REACTOR FACILITY DIRECTOR The Reactor Facility Director (RFD) is the senior lleensed operator responsible for administration and operation of the Reactor Facility and for determination of applicability of procedures, experiment authorizations, and maintenance operations. The RFDs " designee", a person appointed orally or in writing by the RFD and meeting the requirements of fection 0.1.3.1, may perform any functions of the RFD in his absence.
Safety Analysis: This is an administrative change to clarify an existing statement, There are no safety implications, s
- 5. Pares 3-4, Sections 1.17 to 1.27: Adjust nmabering of sections to reflect the addition of Section 1.17 above. :
Safety Analysis: There are no safety implications, t O. Ence 7/P. Section 3.1.?: Replace the Basis in its entirety, whc - 1) "model of the AFRRl TRIGA reactor" is replaced with "Model of the ALRI- TRIGA Reactor" and 2) "instumentated" is replaced with " instrumented" for grammatical correction, as follows:
Based upon the Fuchs-Nordheim mathematical model (cited by C.E.
Clifford et al. in the April 1901 GA Report #2119. "Model of the AFRRl-TRIGA Reactor"), an insertion of 2.8% Ak/h results in a maximum average fuel temperature of les' than 550' C. thereby staying i
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. within the limiting safety settings that protect the safety limit. The 50' C margin to the Limiting Safety System Setting and the 450' C margin to the safety lircit amply allow for uncertainties due to extrapolation of measured data, accuracy of measured data, and location of instrumented fuel elements in the core.
Safety Analysis: This is an administrative change to correct spelling and style. There are no safety implications.
- 7. Pace 8. Section 3.1.3: Replace " controls rods" with " control rods" in the first sentence of the Applicability. Replace " conditions of operations" with
" condition of operation" in the last sentence of the Specifications.
Safety Analysis: These are administrative changes to correct spelling or style. There are no safety implications.
- 8. Pare 10. Section 3.2.2. Table 2: Add an additional minimum reactor safety system scram channel to this Table 2 as follows:
TABLE 2. h11NIh!Uhi REACTOR SAFETY SYSTEh! SCRAhiS hiaximum hiinimum Number in hiode Channel Set Point Steady State Pulse Fuel Temperature 600'C 2 2 Percent Power, High Flux 1.1 hiW 2 0 Console hianual Scram Bar Closure switches 1 1 High Voltage Loss to Safety Channels 20% loss 2 1 Pulse Time 15 seconds 0 1 Emergency Stop (1 each exposure room, 1 on console) Closure switch 1 1 Pool Water Level 14 feet from top of core 1 1 Watchdog (DAC to CSC) On digital console 1 1 Safety Analysis: This change is to reflect USNRC's request to add Watchdog sram requirement for new reactor instrument ation.
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. 9. Parc 10/11 Section 3.2.2: Replace the Basis in its entirety where " safety chambers" replaces " safety channels" to clarify requirement and to include the basis for the watchdog scram requirement.
The fuel temperature t.nd power level scrams provide protection to assure that the reactor can be shut down before the safety limit on the fuel element temperature will be exceeded. The manual scram allows the operator to shut down the system at any time if an unsafe or abnormal condition occurs. In the event of failure 'of the power supply for the safety channels, operation of the reactor without adequate instrumentation is prevented. The preset timer innutes that the reactor power level will reduce to a low level after pulsing.
The emergency stop allows personnel trapped in a potentially hazardous exposure room or the reactor operator to stop actions through the interlock system. The pool water level insures that a loss of biological shleiding would result in a reactor shutdown. The watchdog scram will insure adequate communication between the Data Acquisition and Control (DAC) and Control System Computer (CSC) units."
Safety Analysis: This change is to reflect USNRC's request to add requirement for new reactor instrumentation.
- 10. Pace 22. Section 4.2.5: Replace the Specifications in its entirety as follows to clarify the requirement for fuel element surveillance and to accommodate the fuel follower control rods:
All the fuel elements present in the reactor core, to include fuel follower control rods, shall be inspected for damage or deterioration, and measured for length and how at intervals separated by not more than 500 pulses of insertion greater than $2.00 or annually (not to exceed -15 months),
whichever occur first. Fuel elements in long-term storage need not be measured until returned to core; however fuel elements routinely moved to temporary storage shall be measured every 500 pulses of insertion greater than (2.00 or annually (not to exceed 15 months), whichever occurs first.
Safety Analysis: Fuel in long-term storage is not subject to the rigors of the fuel used in the reactor core. Thus damage to the fuel which may manifest itself as. elongation or lateral bow is not a possibility for fuel in long-term storage. The high degree of purity maintained in the AFRRI TRIGA pool assures that deterioration or corrosion of the cladding will not be a problem.
- 11. Pace 22, Section 4.2.5: Replace the Basis in its entirety for clarification as follows:
The frequency of inspection and measurement is based on the parameters most
.likely to affect the fuel cladding of a pulse reactor, and the utilization of fuel elements whose characteristics are well known.
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. The limit of transverse bend has been shown to result in no diffic tity in disassembling the core. Analysis of a worst case scenario in which twe adjacent fuel elements suffer sufficiently severe transverse bends to result in the touching of the fuel elements has shown that no damage to the fuel elements will result via a hot spot or any other known mechanism.
Safety Analysis: This is an administrative change in the wording to clarify the passage. There are no safety implications.
- 12. Pare 23. Section 4.4: Replace the Specifications in its entirety as follows to conform with latitude recommended in ANSI 15.4/7:
Old Snecifications:
P The operating mechanism of the positive scaling dampers in the reactor room '
ventilation system shall be verified to be operable and visually inspected at least monthly.
- New Specifications:
The operriting mechanism of the positive scaling dampers in the reactor room '
ventilation system shall be verified to be operable and visually inspected at least monthly, not to exceed six weeks. ,
Safety Analysis: This is to conform with latitude recommended in ANSI 15.4/7.
- 13. Parea 25. Section 5.2.1: Replace the Applicability to accommodate fuel follower control rods with the following:
Aonlicnhilitv These specifications apply to the fuel elements, to include fuel follower control rods, used in the reactor core.
Safety Analysis: See attached safety analysis on FFCR-.
- 14. Pace 20. Section 5.2.1: Replace the part "a." of the Specifications in its entirety to accommodate fuel follower control rods with the following:
- a. Uranium content: Maximum of 0.0 weight percent enriched to less than 20% uranium-235. In the fuel follower, the maximum uranium content will '
i- be 12.0 weight percent enriched to less than 20% uranium 235, Safety Analysis: See attached safety analysis on FFCR.
- 15. Parc 26. Section 5.2.1: Replace the Basis in its entirety to accommodate fuel follower control rods with the following:
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' A maximum uranium content of 9 weight percent in a standard TRIGA clement is greater than the design value of B.5 weight percent, and .
encompasses the maximum probable variation in individual elements. Such an l Increase in loading would result in an increase in power density of leds than OST. An increase in local power density of 0% in individual fuel element !
reduces the safety margin by 10% at most. The hydrogen to zirconium ratio of 1.7 will produce a maximum pressure within the cladding well .below the rupture strength of the cladding.
i The power density of a 12.0 weight percent fuel follower element of the same ;
diameter as a control rod, which is smaller than the standard TRIGA element, I will produce the sarne power density in the local area as the standard 8.5 ;
weight percent TRIGA clements due to its increased hydraulle diameter. !
Safety Analysis: See attached safety analysis on FFCR.
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- 10. Pare 27. Section ' 5.2.2(ei: Replaec ".0025 inch" with *0.0025 inch". !
4 Safety Analysis: This is an administrative change to correct style. There are no safety implication 6. j
- 17. Pare 27. Section 5.2.3: Replace the part "a." of.the Specifications in its 4 entirety to accommodate fuel follower control rods with the following: l a.The standard control rods shall have scram capability, and shall contain borated graphite, 04C powder, or boron and its compounds in solid form as l a poison in aluminum or stainless steel cladding. These rods may have an :
aluminutn, air, or fuel follower. If fuel followed, the fuel region will conform to the SpeciDentions of 5.2.1." 3 Safety Analysis: See attached safety analysis on FFCR.
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- 18. Pare 27. Section 5.2.3: Replace " Scram capabilities are provided for..." with
" Scram capabilities are provided by the ..." in the fourth sentence of the Basis.
Safety Analysis: This is an administrative change to correct style. There are no safety implications. ;
- 10. Pane 25. Section 5.3: Replace the first sentence of the Basis with "The limits imposed by this specification are conservative and assure safe storage and !
handling.", .
Safety Analysis: This is an administrative change to correct grammar, f There are no safety implientions.
- 20. Pene 29. Section 0.1.1: Replace " Chairman, Radiation Sciences _ Department"
- with " Chairman, Radiation Sources Depart!nent to administratively reflect new Page 15
. name. Replace "... Control Chain, in whieh" with *... Control Chain in which..."
in last sentence of the section.
Safety Analysis: These are administrative changes to clarify existing statements and correct grammar. There are no safety implications.
- 21. Pare 20. Firure 1: Replace " Chairman, Radiation Safety Department" and
" Chairman, Radiation Sciences Department" with " Chairman, Safety and Health Dept." and " Chairman, Radiation Sources Dept " respectively.
Safety Analysis: These are administrative changes to clarify existing statements. There are no safety implications.
- 22. Pare 30. Section 0.1.2: Change the Responsibility in its entirety to clarify the responsibility of the Reactor Facility Director with the following:
The Director. AFRRI. shall have license responsibility for the reactor facility.
The Reactor Facility Director (RFD) shall be responsible for administration and operation of the Reactor Facility and for determination of applicability of procedures, experiment authorizations, maintenance, and operations. The RFD may designate an individual who meets the requirements of Sntion 0.1.3.a to discharge the RFD's responsibilities in the RFD's absence. During brief absences of the Reactor Facility Director and his designee, the Reactor Operations Supervisor shall discharge these responsibilities.
Safety Analysis: This is an administrative change to clarify existing statements, correct grammar, spelling, or punctuations. There are no safety implications.
- 23. Pace 30. Section 0.1.3.1: Replace the requirements of Reactor Operations Supervisor (ROS) in its entirety with the following:
- b. Reactor Operations Supervisor (ROS)
At the time of appointment to this position, the ROS shall have 3 years nuclear experience. Higher education in a science or nuclear engineering field may fulfill up to 2 years of experience on a one-for-one basis. The ROS shall hold a USNRC Senior Reactor Operator license on the AFRRI reactor. In addition, the ROS shall have 1 year of experience at AFRRI or -at a similar facility before the appointment to this position.
Justification: A year of experience at AFRR1 or at a similar facility provides an adequate experience for a person to assurne ROS duties at AFRRI.
- 24. Pare 31. Section 0.1.3.2(cH2h Replace Radiation Safety Department" with "8afety and llealth Department".
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. Safety Analysis: This is an administrative change to show a change in -
existing tcrminology. There are no safety implications.
- 25. 'are 31. Section 0.2.1.1(al: Change from
- Chairman, Radiation Safety Jepartment, AFRR1". to
- Chairman, Safety and Health Department, AFRRI" t to administratively reflect new name.
1 Safety _ Analysis: This is an administrative change to show a change in ,
terminology. There are no safety implications. +
- 20. Pace 32. Section 0.2.3.2: Change from " semi annually" to " semi annually" for typographical correction. -
1 Safety Analysis: This is an administrative change to correct grammar.
There are no safety implications. ;
- 27. Pare 34. Section 0.3.1: Replace "The activities are as follow:" by "The .
activities are as followr:" to correct grammar.
Safety Analysis: This is an administrative change to correct grammar, i There are no safety implications.
- 28. Pare 34. Section 0.4.1: Change from " Reactor Facility Director, Reactor Branch" to " Reactor Facility Director" and from " Radiation Safety '
Department" to " Safety and Health Department" to administratively reflect dew Dameb.
Safety Analysis: This is an administrative change to reflect reorganization.
The change was approved by the RRFSC meeting held on 25 July 1989. A copy of the chenge was sent to the USNRC on 28 July 1989,
- 20. Pare 35. Section 6.4.3: Replace "be" with "by" for grammatical-correction in the second sentence.
Safety Analysis: This is an administrative change to correct grammar.
There are no safety implications.
- 30. Pace 35. Section 0.5.1 c.: Replace "be" with "by" for grammatical correction in the second sentence.
- Safety Analysis: - This is an administrative change to correct grammar.
There are no safety implications.
- 31. Pace 35. Section 0.5.1 d.: Replace "to NRC' with "to the NRC" for grammatical correction.
Safety Analysis: This is an administrative change to correct grammar.
There are no. safety implications.
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