ML20125C565
| ML20125C565 | |
| Person / Time | |
|---|---|
| Site: | Oregon State University |
| Issue date: | 12/18/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20125C563 | List: |
| References | |
| NUDOCS 8001100414 | |
| Download: ML20125C565 (9) | |
Text
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l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 4 TO FACILITY OPERATING LICENSE NO. R-106 =
DOCKET NJ. 50-243 OREGON STATE UNIVERSITY I. Introduction .
By letter dated April 16, 1979, supplemented by letters dated July 11, '
August 17 and October 10, 1979, the Oregon State University (OSU or the licensee) requested amendment to Facility Operating License No. R-106 for the OSU TRIGA keactor (OSTR). The amendment would provide sixteen (16) i changes to the Technical Specifications (TS) grouped as follows: '
! (A) Proposed Changes Nos. I through 4 relate to a proposed upgrading of the reactor control console.
(B) Proposed Changes Nos. 5 through 12 relate to proposed new limits on core configuration for an operational core, a proposed increase in i allowable reactivity insertion for pulsing, and a proposed increase j in reactivity worth of any single experiment.
I
! (C) Proposed Changes Nos.13 and 14 are proposed changes to the Administra-
! tive Section of the TS which would reflect a change in the licensee's organization.
(D) Proposed Change No.15 would change the calibration frequency of the l fuel temperature channels from semiannual to an annual basis.
(E) Proposed Change No.16 would extend the time period for submitting the annual report from 60 days to 75 days following the 30th of June of each year.
II. Discussion and Evaluation Each of the above items are discussed and evaluated separately below.
(A) Proposed Chances Nos.1 throuch 4 relatina to uoarading the reactor control console D4scuss o" 90008240 The present OSTR console was installed in 1967 and contains many printed circuit boards which are no longer available. The licensee does not have a complete set of spare boards and is concerned that the reactor may be shut-down for an unreasonable length of time if problems with the console elec.
8001 100M
tronics are experienced. Therefore, the licensee has decided to purchase from General Atomic, the console manufacturer, new electronic packages to upgrade their TRIGA console. These packages are standard instrumentation on new TRIGA consoles.
The proposed TS changes to Sections 3.5.2, " Reactor Control System" and 5.5.3, " Reactor Safety System" are needed to reflect the addition of this new instrumentation. The new instrumentation, which utilizes all solid state modular construction with integrated circuitry, would provide increased reliability over the existing instrumentation and, therefore, upgrade the console electronics.
The proposed new instrumentation package includes:
(a) A 9.5 - decade multirange' linear channel (Model NML-2)
(b) A 10 - decade log power channel (Model NLW-2) -
(c) A period circuit (Model NR-4)
(d) A linear safety channel (Model NP-5)
(e) A preamplifier (Model PA-5)
(f) Pulsing logic (g) Calibration circuits for linear and log power and period, and (h) Power supplies, including a high voltage supply (Model HV-6).
The relationship between the present instrumentation and the new instrumentation is shown in Figure 1.
I Eval ua tion The proposed modifications to the console instrurentation consist of:
(1) Replacinc the cresent multirance linear char.nel usinc an ion enameer with the new 9.5 decade linear enanr.el criven by a fission chamber. Inis same fission enameer is also USEd to drive tne new 10-decade log channel. We view tnis arrange-ment as a single " linear-log" channel, as failure of the single detector (fission chamber) means the loss of both linear and log information to the operator.
Since this new fission chamber is physically larger than the existing detector , it cannot be located in -he same position.
The new fission chamber would be placed int: the existing log ion chamber shrout;, which would accept the :hysical size of the new chamber . We have reviewed this arrangement and find that the new location would basically provide the same source-fuel-detector geometry as the existing detector, and would not constitute a problem with detector shadowing. We find this change in detector location acceptable.'
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3-The present linear channel performs both safety (scram) and control functions. The linear portion of the new " linear-log" channel would provide a signal to the servo system for auto-matic power level adjustment. The new safety power level =
channel would provide the scram function.
(2) Replacina the cresent multirange loc channel with a ceriod circuit us1no an ion enameer with a new 10-cecaoe loc cnannel, also witn a ceriod circuit, criven by a fission cnaeber. This new instrumentation is considered to be the log portion of the new " linear-log" channel. This channel would provide the start-up interlock function (preventing control roc withdrawal at a count rate of less than 2 cps) currently perforrec by -he count-Pate (startup) channel. The count rate enannel would be -
no longer required.
The new calibration circuits for the log and linear oower and oeriod circuits are similar to -he existing calibratien circuits ;
in that they generate test signals to the enannel electronics I for checking proper circuit alignment. Six different calibra-tion signals are provided for calibration of both of the log and linear circuits. Two separate period calibra-ion signals are used. The new calibrate switch (period / log test switch) e is not spring-loaded as are the existing switches. To preclude leaving the calibrate switch in a calibrate position, the switch would be connected to the source and 1Kw interlocks. We find this arrangement to be acceptable.
(3) Addinc a new " safety cower level" channel, with scram cacability, driven bv an ion enameer. Inis enannel woulc be identical (except for new electronics) to the present percent power channel.
The two channels would use separate ion chambers. This safety power level channel would provide scram capability. This channel, however, would trip only at 110% of full power (i.e. ,1.1 MW) whereas the present linear power channel trips at 110% of each range. Block diagrams of the present and proposed design are shown in Figure 1.
The use of the proposed wide-range " log-linear" channel with the automatic servo presents a different situation regarding the separation of safety and control instrumentation.
The existing instrumentation cannot cause the loss of both auto-matic power level adjustment and period protection via a single detector failure. Upon loss of the control signal to the servo system, period protection is still provided for a reactor trip, in addition to the fuel elemen temperature trip and a 100%
neutron level trip. However, a period-limiting circuit in the existing design limits the regulating rod speed so that the period never gets shorter than about 12-15 seconds, thus period scram protection (period <3 sec.) is not utilized when in the 90008242
automatic mode. The licensee has stated that the maximum increase ;n fuel temperature, before scram via a level trip, would be about 240'C from ambient or about 260*C. This temp-erature is well below the limiting safety system setting (LSSS) for fuel temperature (510*C), which in itself incorporates a large safety margin before the fuel temperature safety limit (1150 C) is reached.
Failure of the fission chamber detector in the proposed system causes the loss of both period protection and the control signal-to the servo system for automatic power level adjustment. Pro-tection would still be provided by two neutron level trips at 110% of full power and a fuel element temperature tr'p at 510*C. The licensee has stated that the maximum increase in fuel tempature resulting from this detector failure before a reactor trip (via one of the two level trips) would be about 40*C from ambient, or approximately 60*C. -
Therefore, even though period protection would be lost due to failure of the fission chamber in the new design, the increase in fuel temperature would actually be less than that for a similar detector failure in the present system. In addition, the reactivity insertion rates postulated above are not nearly as rapid as during a routine pulse for which an acceptable safety analysis has been documented.
For the above reasons, we find the deletion of level trips at 100% of each range and the period / control circuitry configurations to be acceptable.
(4) Removino the count-rate (startuo) channel. Its interlock function, which is to prevent control rod withdrawal at count rates less than 2 cps, will be taken over by the new " linear-log" channel.
The licensee has determined that overpower conditions will not produce saturation or fold-over in any of the proposed new instrumentation channels.
All minimum reactor safety channel functions, interlock functions, and operable measuring channels required by the current TS will remain unchanged.
We have reviewed the proposed modifications to the OSTR console instrumentation described above, and find these equipment and design modifications acceptable and would not reduce the margin of safety. We have also reviewed the other console electronics included in the proposed package (i.e., pulsing logic, calibration circuits, preamplifier, and power supplies), and have found this instrumentation to be acceptable, (5) Technical Soecification (TS) chances The following proposed changes to the OSTR TS are associated with the modi-fications to the console electronics:
1.
In Section 3.5.2 (Reactor Control System), in the table listed in the Speci-fication:
Add " Safety Power Level" as a measuring channel, effective in the steady-state (s.s.) and square-wave (s.w.) modes.
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- 2. In Section 3.5. 3 (Reactor Safety System), Table I:
Change " Log Power Level" in column 1.to read " Wide-Range Log Power Level."
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- 3. In Section 3.5.3 (Reactor Safety System), Table I:
Change " Linear Power Level" in column 1 to read " Safety Power Level."
- 4. In Section 3.5.3 (Reactor Safety System), Table II:
Change " Count-rate Channel" in column 1 to read " Wide-Range Log Power Level Channel ."
We find these proposed modifications to Sections 3.5.2 and 3.5.3 of the TS .
adequately reflect the proposed changes in console electronics for the OSTR and are correct, and, therefore, are acceptable.
In summary, based on our review of tne licensee's submittal, we conclude that the proposed modifications to the OSTR console electronics and the associated TS cnanges are acceptable, wculd not reduce the r.argin of safety, and would not increase tne probability or ecnsecuences of an accident.
(B) Procosed Cnanges Nos. 5 throuch 12 relating to limits or the core conficu-ration, tne reactivity insertion for culs1ne, anc :ne react 1vity wor:n of
- "any sinale exoeriment Discussion l
The present limits on the core configuration and operation were initially '
supported by the licensee's Safety' Analysis Report (SAR) dated April 3,1975, as revised September 11, 1975. The operating limits were establisned on the limiting core configuration by the SAR such that ouising would produce pulse transients with maximum fuel temperatures no greater than 950 C in the FLIP fuel and 800*C in the standard fuel; i.e., a safety margin of 200*C with respect to the safety limits of the fuel.
The licensee established, in August 1976, an operational core consisting of 85 FLIP fuel elements (a full FLIP core). The "Startup Report for the Full FLIP Fuel Loading" dated May 30, 1977, provided data that not only con-servatively confirmed the analyses of tne SAR, but also verified that pulsing could be increased to 2.60 dollars on a full FLIP core, and the pulsing would produce pulse transients with maximum fuel temperatures no greater than 950*C in FLIP fuel and 800*C in standard fuel if it were in the outer region of the core.
The licensee's request would: (1) increase the minimum number of FLIP elements from 56 to 80 in a contiguous block in the central region of the core, (2)'
increase the allowable reactivity insertion for pulsing from 2.35 to 2.55 dollars, and (3) increase the allowable reactivity worth for a single experiment from 2.35 to 2.55 dollars.
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i D D The licensee has indicated that such a proposed core. configuration would behave essentially as a full FLIP core in operation. The licensee based this upon the analysis in the SAR which concluded that the maximum power per fuel element in a core with 55 FLIP fuel elements in a contiguous block in the central region of the core would differ little from the maximum oower per fuel eletent in a full FLIP core. Therefore, the difference in maximum power per element between,the proposed core configuration and a full FLIP core is not significant. Hence, the proposed core configuration would .
behave essentially as full FLIP core and an increase in reactivity insertion to 2.55 dollars would produce pulse transients with maximum fuel temperatures no greater than 950*C for FLIP fuel elements and 300*C for standard fuel elements.
The licensee's basis for the proposed change of the reactivity worth of a single experiment to 2.55 is that it can be the same as the pulsing limit on reactivity.
TRIGA Standard and TR:GA FLIP fuel have distinctive markings on the upper
- ip of each fuel elemen:. Fuel loading procedures use these markings to assure the procer positionine of each fuel element in the core a 2 therefore assure tnat a stahdard element would only be placed in the outer region of the core where power levels are the lowest.
Evaluation We agree with tne licensee that the proposed operational core would behave essentially as a fu'.1 FLIP core. Based on the data of the Startup Report
- ated May 30, 1977, we agree that the proposed operational core would produce pulse transients with maximum fuel temperatures no greater than 550 C in the FLIP fuel and 800*C in the standard fuel when pulsed with reactivity insertion no greater than 2.55 dollars. This would maintain a safety margin of 200*C with respect to the safety limit of the fuel. We agree with the licensee that the limit on reactivity worth of a single experi-cent can be the same as the limit on reactivity insertion for pulsing.
Therefore, we find the licensee's proposed operational core configuration,
- roposed limit on reactivity insertion for pulsing and proposed reactivity north of a single exoeriment to be acceptable and would not reduce the rargin of safety and would not increase the orobabilitv or consequences of an accident. .
90008245
7-5
(:) Proposed Changes Nos.13 and la relatino to caanaes in the licensee's organization Ciscussion -
This proposed chances reflect a change of title of one of the licensee's principal officers The office of the former Dean of Administration has been changed to Vice President for Administration. The Reactor Administrator would be responsible to this Vice President for the safe operation and raintenance of the reactor.
Evaluat'on i
We have reviewed the licensee's proposed changes and find them to be administrative in nature, acceptable and would have no affect on reactor .
safety.
(D) Proposed Change No.15 relatine to calibration frecuecy of fuel temo-erature cnannels We have discussed with the licensee their propcsed change in calibraticn frequency of the fuel temperature cnannels from a semiannual to an annual basis. As a result of our discussions, the licensee has agreed to withdraw the recuested change in
. calibration frequency and to continue with the current requirements of the TS.
'E) Procosed Chance No.16 relatinc to a chance in the time for suomittine the annual recor:
Discussion The licensee has reouested an additional 15 cays in which to submit the annual report. The change woulc provide approxi-mately one month to precare the report after all the data are available.
Evaluation We have review?d the licensee's request and nave determined that extending the time for submitting the annual report by 15 days is not significant and therefore is acceptable.
II;, Environmental Consiceration We have determined that this amendrent will not result in any significant environmental impact and that it does not constitute a major Commission action sigr.ificantly affecting the quality of the human environment. We have also determined that this action is not one of those covered by 10 CFR 5 51.5(a) or (b). Having made these determinations, we have further concluded that, pursuant to 10 CFR i 51.5(d)(4), an environnental impact statement or environmental impact appraisal and negative declaration need not be pre-pared in connection witn issuance of this amendment.
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IV, Conclusion '
We,have concluded, based on the considerations discussed above, that: '
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed mancer, and (3) such activities will be conducted in compliance with the Comission's regulations and the issuance of this amendment will not be inimical to thethe common defense and security or to the health and safety of public. -
Dated: December 18, 1979 90008247
=
- - Flux Regulator I
' ~ Flux Regulator
- L I ""d
Linear Channel ---*- Recorder CIC -*- Hecorder Channel l _ Scram Logic FC -
g,
= Startup Interlock to (110% of each range) . __ L poise interiock (igu L g Channel *- Recorder
- -- Recorder CIC Log Channel
- > Pulse Interion.k I g (i tw} gogic Period Circuit .- Scram Lo9i c Period Circuit *- Sc ram log ic -
(0 3 sec.)
UIC v Linear Percent = Scram logic Power Channel (110% of full Power)
UIC Linear Percent Scram logic g,ower Channel (110% of full power)
Safety %g Power Channel (110% of full power)
Startup Startup Interlock FC Channel ,lo9IC w
O O PRESINT ARRAllGEllENT PROPOSED DESIGN O
CD rv FIGURC 1 b
CD I
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