ML20116J238

Annual Rept of Changes,Tests & Experiments Performed Under Provisions of 10CFR50.59 for Oregon State University Triga Reactor (Ostr)

ML20116J238
Person / Time
Site:	Oregon State University
Issue date:	06/30/1992
From:	Andrea Johnson Oregon State University, CORVALLIS, OR
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9211160102
Download: ML20116J238 (28)

Text

v MiAdoNbNIER 1

November 9,1992 U. S. Nuclear Regulatory Commission ATTENTION: Document Control Desk OREGON Washington, D.C. 20555 STATE UNIVERSITY

Subject:

Annual Report of Changes, Tests and Experiments Performed Under the Provisions of 10 CFR 50.59 for the Oregon State University TR'GA Reactor (OSTR), License No. R-106, Docket Radiationrenter A100 No. 50 243.

Corvams, Oregon 97331 5903 The following report is submitted in accordance with the requi ements of 10 CFR 50.59(b) and 10 CFR 50.4, and cover.s the OSTR's annual reporting -

period of July 1,1991 through June 30,1992, The informotion in this -

report is compiled annually and is submitted to the USNRC in this specific 10 CFR 50.59(b) report, as well as in a special section of the OSTR annual' report, which was submitted on October 29,1992.

During the specified reporting period there were eight change:: to the reactor facility and four changes to reactor procedures conducted pursuant to 10 CFR 50.59. There were three changes to reactnr experiments, no tests, and no new experiments performed under the provisions of 10 CFR 50.59 during the current reporting period, The individual changes being reported are listed below by category and by title, and are described in more detail in ' Attachment A~. Regarding this attachment, you wili nota that it includes a brief description of each change followed by a summary of the safety evaluation conducted for the described :

Telephonc change. As required, none of the changes performed under the provisions .

So3. m. nu of 10 CFR 50.59 necessitated a change in the .OSTR Technical Specifica-

~

tions or involved an unroviewed safety question as defined in 10 CFR ru - 50.59(a)(2).

503 m .0480 l

130073 9211160102 920630 gDR- ADOCK 0500 3

'b 1

• . s.

t

.. USNRC November 9,1992

1. Chanaes to the Reactor Fac"itv:

a. Addition of Ventilation Ducting for Reactor Bay Sample Decapsulation Hood

b. R= .nlacement of the GM Tube in the Reactor Water Radioactivity Monitor

c. Replacement of the Reactor Water Radioactivity Monitor

d. Replacement of the Primary Water Radioactivity Monitor (Revision)

e. Modification to the Reactor Console Left Hand Drawer, Fuel Element Temperature Monitoring System, and to the Thermocouple Calibration Procedures for Fuel Temperature and Reactor Tank Wster Temperature

f. Replacement of an Air Flow Gauge on the Stack Effluent Monitor

g. Replacement of the Isokinetic Sampling Probe on the Stack Effluent Monitor

h. Automatic Shut-off of the D102 Hood Fan

2. 10 CFR 50.59 Chances to Reactor Procedures a, Revisions to the OSU Radiation Center and TRIGA Reactor. Emergency Response Plan and Emergency Response implementing Prococures

b. Change of Time interval Between Transient Rod Calib.ations

c. Minor Procedurbi Revisions to OSTROP 26, " Procedures for the-Use of External Monitoring and Recording Devices"

d. Temocrary Procedural Addition to OSTROP 26, " Procedures for the Use of External Monitoring and Recording Devices"

3. 10 CFR 50.59 Chances to Reactor Exoeriments

a. Revision of Reactor Experiment B 23

b. Revision of Reactor Experiment B-30

c. Second 1991-92 Revision of Reactor Experiment B-23

L y, (:- .1 . ,

p ._ _. . , .

.., .USNRC- ' 3-

- . Nov:mber 9,1992

• ~ '

. 5 We trust thet you wiil find this year's report to be in good order. However, should you require s

-more info'rmation or have questions regarding our report, please let me know.-

Yours sincerely,.:  ;

A.l .J hnson -,

Dir tor i Radiation Center ,

AGJ:Jrs\rc\5059tept.let Enclosure cc: Regional Adrainistrator, Region V, USNRC, Walnut Creek, California ,

Mr. Lero) Norderhaug, Region V, USNRC, Walnut Creek, California .

Mr. Phil Cualls, Region V, USNRC, Walnut Creek, California '

Mr. Al A<*ims OSTR Project Manager; USNRC; Washington, D.C.

Mr. DaviSStewart-Smith, Oregon Department of Energy, Salem,' Oregon- ~

T.V.Anu rson, Reactor Supervisor, OSTR S. E. Bint /, Chairman, Reactor Operations Committee, OSTR B. Dodd, Reactor Administrator, OSTR J. F. Higginbothaml Senior Health Physicist, OSTR l 4

(

3 g

• 4 l

l

$ .(  %

, , >+ .n

• s ATTACHMENT A Chanaes to the OSTR Facility, to Reactor Procedures, and la Rc.astor Exogriments Performed Pursuant to 10 CFR 50.59 The information contained in this section of the report provides a summary of the changes performed during the reporting period under the provisions of 10 CFR 50.59. For each item listed, we have included a brief description of the action taken and a su.nmary of the applicable safety evaluation. Although it may not be specifically stated in each of the following safety evaluations, all actions take i under 10 CFR 50.59 were implersnted only-af ter it was established by the OSTR Reactor Operations Committee (ROC) that the proposed activity did not require a change in the facility's Technical Specifications and did not introduce or create an unreviewed safety question as defined in 10 CFR 50.59(a)(2).

1. 10 CFR 50.59 Channes to the Reactor Facility There were eight changes to the reactor facility which were reviewed, approved, and performed under the provisions of 10 CFR 50.59 during the reporting period,

a. ADDITION OF VENTILATION DUCTING FOR REACTOR BAY SAMPLE DECAPSULATION HOOD (1) Description In order to provide better air flow through the reactor bay hood which is used to remove sample _ capsules from TRIGA irradiation tubes, the Health Physics staff and the OSU Physical Plant upgraded the ducting leading from the hood to the reactor bay exhaust system. Also, a larger absolute filter was added above the hood. The new ducting rises directly up the wallin the southeast corner of the reactor bay and takes a 45 angle to avoid the third floor windows and 4th floor door. The duct then penetrates the 4th floor wall and runs in the space between the reactor bay intake and exhaust ducts in D400 before entering the exhaust plenum.

a

,4 2-(2) Safety Evaluation Measurements and engineering analysis by Physical P! ant and Radiation Center Staff indicated that the air flow tk ough the hood was significantly improved as a result of the installation of the new ducting. This increased safety by providing more positive hood ventilation and better filtration than before. Thoroforc, in the event of a release of radioactive material in the hood, there will be negligible release to the reactor bay and the environment, and less likelihood of radioactive materialinhalation by persons working at or near the hood, in addition,it was expected that there would be no effect on the total flow rates in and out of the reactor bay, and that the negative differential pressure in the bay would not be rneasurably changed. After the duct work was installed, the differential air pressure was checked and it was confirmed that it was within the normal limits. Air flow measurements were also made in the reactor bay stack which confirmed that the flow cal lbration of the stack monitor was still correct,

b. REPLACEMENT OF THE GM TUBE IN THE REACTOR WATER RADIOACTIVITY MONITOR (1) Description The response of the reactor water radioactivity monitor had become erratic during normal reactcr operations.- The cause of the erratic response was believed to be due to an aging GM detector in the monitoring system. Since .

the GM tube being used was no longer manufactured, the Scientific Instrument Technician proposed to replace the existing GM tube with a new GM tube which had similar response characteristics. This change also necessitated manufacturing a slightly different tube shroud so that the GM tube could be correctly inserted into the center of the water monitor tank.

t .

3-4 (2) Safety Evaluation

~

The new GM tube was selected to have response specifications which were as close as possible-to the old tube, in addition, the new tube was calibrated after insertion to establish its response chnacteristics prior to actual use. Use of the new GM tube should have simply restored the water monitor's responso and stability and thus improved safety.

c. REPLACEMENT OF THE REACTOR WATER RADIOACTIVITY MONITOR (1) Description The GM tube used in the reactor water radioactivity monitor had aged to the point whero it needed replacing; however, an exact replacement was not available. Therefore, the previous safety evaluation (item b) was written to allow a new GM tube with similar characteristics to be installed in the existing system. When the new tube was installed and tested,it was foun<l to be insufficiently sensitive for operational needs. This was partially due to the fact that the console voltage supply to the GM tube was fixed at a value which did not ideally match the voltage requirements of the new tube. As a result, the old original GM tube was returned to service until a different solution could be found.

The Scientific Instrument Technician proposed using an Eberline RM-16 power supply and ratemeter to replace the existing water radioactivity ammeter, which in the process would allow use cf the Eberline ratemeter's power supply to provide a high voltage of 900 VDC for the new GM tube (LND 725). This configuration had been bench tested and calibrated in the range up to 150 mR/h. The water high activity annunciator would be connected to the ratemeter and set to alarm at 50 mR/h. Finally, the old water activity monitor and wiring would be removed. ,

(2) Safety. Evaluation The proposed facility chang would restore the water radioactivity monitor to full, functional reliability. Sensitivity would be approximately equal to or

+ .

4

. 4 greater than that provided by the existing system. - No electrical safety implications for the roccior console were involved with this change. The main impact of the proposed change would be to improve system stability, eliminate falso alarms and thereby increase reactor safety by maintaining operator alertness to the h!gh water activity alarm.

Based on the OSTR's past and current modes of operation, the presence of high radioactivity levels in the reactor's primary water is associated with situations where it is expected that the reactor top continuous air monitor, will usually alarm first, or in conjunction with any water radioactivity alarm, thereby providing extra assurance of detection.

d. REPLACEMENT OF THE PRIMARY WATER R ADIO ACTIVITY MONITOR (R EVISIOE)

(1) Description 10 CFR 50.59 safety evaluations included as items b and c in this report discussed various aspects of the replacement of the primary water radioactivity monitor. This evaluation is a revision of item c and was needed -

because a different approach was ultimately used .to replace certain components in the water radioactivity monitor.

The first change involved the use of a Tracerlab linear ratemeter und pre-amplifier instead of the Eberline RM-16 ratemeter which was located in the beam port #3 area. The Tracerlab ratemeter was already housed in the right .

hand console side cabinet next to the ratemeters for the stack gas and stack ~

particulate monitors, and was part of the original equipment purchased for the OSTR. However, this ratemeter had never been used operationally. The ratemeter was part of e water monitoring system which was not installed, and was originally oquipped with a shielued detector containing a built-in pre'-

amplifier. The shielded detector was still not used, but the pre-amplifier was'-

removed from the detector module and put into a separate box. This box

-was located near the new GM tubo discussed in the previous 10 CFR 50.59 evaluations (items t. and c).

- , e w,

-.5 As a result of the preceding changes, the new configuration for the waterc

' monitoring systern consists of the.new GM tube (LND 725) located inside a machir$ed holder which fits in the center of the water monitoring chamber in --

the demineralizer system. Th 'M tube is attached to the re-amplifier box,--

which in turn is connected to i a Tracerlab ratemeter in the control room.

This system was checked and found to be fully functional and suitable for the indicated purpose.

s An additional change to the previous 10 CFR 50.59 evaluations (items' b and .

c) involved system calibr$ition. More specifically, the new system' was calibrated so that the indicated gross count rate could be used to determine--

^

an acoroximate water radioactivity concentration la pCi/ml. T.Ws calibration .

was performed by first putting the new GM detector in'its normal operational location inside tne water chamber in the demineralizer loop af ter the reactor -

was shutdown. The control room Tracerlab ratemeter was 'then used to j obtain a garmia counting rate.(in gross cpm) at a specific time, which was -

correlated from a time standpoint with the gross pCi/mi of gamma emitters in a sample of reactor primary' water. By making this comparison a number -

of times over a wide range of count rates, a calibration curve was obtamed. -

- The alarm point for the new system was set at a count rate which was just -

far enough above.the maximum' count: rate normally encountered-during i routine operation to minimize the potential for falso alarms.

(2) Safety Evaluation ,

The applicable conclusions from ihe previous safety evaluadons are still valid and are incorporated hero b/ reference (10 CFR 50.59 Safety Evaluations: .. '

items b and c). There are no unfavorable safetyimplications associated with .

the change in electronics.4 The Tracerlab ratemeter is similar to those used 1

-t in the stack gas and stack particulate' monitors. These have been in use for i many years, and have proven to be very reliable.

4 The chant,e with. respect to the mathod of calibration is, technically,much better than that previously use'd, and 'therefore increases the' ability to.-

rw 4 9 w 7 F

quick!y estimate the water radioactivity concentration. Since the primary water radioactivity monitor is used strictly to estimate the gross garnma emitting radioactivity in the primary coolant,-indications of abnormal radioactivity in the primary water would be promptly investigated by complete analysis of individual water samples, and by cross checking with other monitoring systems such as the reactor top continuut; air monitor.

This investigative policy represents no change from current radiation safety procedures. However, the water monitor's current calibration is improved -

and this will lead to a small increase in safety.

e. MODIFICATION TO THE REACTOR CONSOLE LEFT HAND DRAWER, FUEL ELEMENTTEMPERATURE MONITORING SYSTEM, ANDTO THETHERMOCOUPLE CAllBRATION PROCEDURES FOR FUEL TEMPERATURE AND REACTOR TANK WATER TEMPERATURE (1) Description The failure of the second of three thermocouples in the instrumented fuel element lef t only one remaining thermocouple. While this met the Technical ,

Specification requirements for reactor oporation, practically it would not have allowed pulsing of the reactor. .The original system for indication of fuel element temperature also had some limitations.

The amplifier in the reactor console's lef t hand drawer did not have are active -

reforence junction compensation. This meant that if the console temperature -

changed, a difference developed between the actual and the indicated fuel l element temperatures. In addition, the design of the fuel temperature input i'

electronics in the left hand drawer was such that it would not accept emf inputs from a standard such as the Portametric Voltmeter Bridge, so that a.

direct calibration method was impossible. Finally, with the existing design, ]

the electronics did not provide the flexibility to operate in ' pulse mode with i only one functional thermocouple in the instrumented element.

d 1

4

l i 7 The Scientific instrument Technician proposed modifications and designed circuits to address the above limitations. This included the installation of a new amplifier in the lef t hand drawer. This amplifier was designed to accept the output of the optional analog output unit card installed in the console's digital temperature indicator. The analog output unit produces an output of 1 mVDC for every 1'C indicated on the digital display. The result of these changes is that the left-hand drawer temperature meter reads the 5,ame thermocouple as that selected on the digital temperature indicator. in eddition, the change enables the reactor to be pulsed because the new electronics provide the necessary fuel temperature input to the console chart recorder even if there is only one operational thermocouple in the instrumented fuel element.

The specific steps involved in the modification were:

(a) The A16 card in the lef ~ hand drawer was s ' aced with the modified card.

(b) A cable was run from the left-hand drawer to the console's digital temperature indicator.

(c) The wiring from the trip test switch was disconnected and the fuel zero switch was removed from the left-hand drawer.

(d) A trimmer pot was installed in place of the fuel zero switch and the voltage divider was reconfigured for the calibrate switch. This allows the operator to check the left-hand drawer temperature circuitry by .

placing the calibrate switch in the catibrate position The trimmer pot was adjusted to provide a full scale meter deflection as part of the alignment procedure.

System and circuit diagrams of the changes are shown in Figures IV.D.1 and IV.D.2.

(2) Conforming Procedural Changes in addition to the facility changes described above, the reactor staff revised the parts of OSTROPs 15 and 16 which relate to how thermocouple i

~8-

~

calibrations are made for fuel element temperature channels and reactor tank water temperature channels. The original procedure involved actually raising the water and fuel temperatures by operating the reactor, noting tha indicated temperatures on the instrumentation, and comparing them with that measured by the Portametric.

A much better method scientifically, as well as procedurally, is to use the Portametric directly as a standard to input a known voltage to the temperature measuring instrumentation. This also eliminates the need to operate the reactor to calibrate the thermocouples.

(3) Safety Evaluation The changes contribute to increased safety by eliminating all . I the previously noted limitations of the original system in the console left hand '

drawer. Safety is further enhanced by the fact that there are two fuel j temperature displays on the console from the selected thermo- couple.

The new A16 card has an indicating function only, so that any failure of this circuit will result in an easily recognizable anomalous reading on the fuel- '

temperature meter in the left hand drawer. The analog output from the digital temperature indicator is isolated so that t is not possible for a failure. "

of the A16 card to cause a problem with the digital unit.

The procedural change has no safety implication, in that the Portainetric is still the standard to which the other instrumentation is compared. The reactor will not now be operating during thermocouple calibration, and,-

. thcrofore, there are-no reactor safety implications from the operational.'

viewpoint.

I b

1

. .g.

f. REPLACEMENT OF AN AIR FLOW GAUGE ON THE STACK EFFLUENT MONITOR (1) Description The stack effluent monitor has two air flow gauges in the system. The health physics staff removed the original flow gauge assembly (which was worn out) from the stack effluent mr...itor and replaced it with a new flow gauge. The now gauge was placed in line with the air flow after the gas-monitoring chamber and before the low flow sensor as shown in Figure IV.D.3. In addition, the existing flo ontrol valve was moved to a more accessible location near the flow gauge . The new gauge (gauge No. 2) was mounted onto the same stand which supports the other existing gauge (gauge No.1).

(2) Safety Evaluation The replacement of the flow gauge has no affect on reactor safety, but increases radiological safety. The purpose of the new flow gauge is to-provide a more accurate method of determining whether or not there is any -

air leaking into the stack monitoring system. Air leaks into.the . system reduce the accuracy of the monitoring data and thus must be prevented.

The logic associated with the two air flow gauges is as follows: Gauge No.

1 measures the air flow rate where the air enters the stack monitor, anJ:

gauge No. 2 measures the flow rate after the air.has-passed through all monitoring chambers, if the air flow rate on both gauges is the same, then there is good as;urance that there are no air leaks'into the- system.

However, if the second gauge reads higher than the first gauge then it is:

likely that air leaks exist. With this latter informstion, leaks can be quickly -

found and repaired. As part of this process, the new gauge will help obtain more accurate air flow readings than the original gauge, which was worn to the point of replacement.

Moving the flow control valve to a more accessible location has no effect on the system, it just makes it easier to adjust the flow. The originallocation

, . _ _~

'- - e 4 10-

+ .

A 6 of the valve was low and down in the middle of the monitor's piping _which-

- made ' adjustment very awkward. .

. g. REPLACEMENT- OF THE ISOKINETIC SAMPLING- PROBE -'ON THE STACK-EFFLllENT MONITOR -

(1) Description  ;

After installation of the new stack monitor air flow gauge described in 'the previous 10 CFR 50.59 safety analysis (item f),it was observed that when - ,

the monitor's air flow rate was set to sample the stack at the required-isokinetic flow rate, the second flow meter located after the particulate and gas monitoring chambers read off scale. Extensive investigations determined :

that the higher reading on the second air flow gauge was ag.1due to leakage t

of air into the monitor, but rather was'due to the pressure drop across 'he -

particu! ate filter paper. It was also determined that this pressure ' drop (and  :

the flow rate difference shown on the two flow gauges) becomes much less --

at flow rates clQhtly lower than the current isokinetic flow rate. Therefore, the reactor operations staff and the health physics staff replaced the stack sampling probe with one of identica' design, but of slightly smaller diameter.'

The new probe reduced the volume flow rate (in CFM) required to maintain ,

isokinetic sampling in the stack, and, at a result, both the intake and outflow ,

air gauges remain on scale. At the new lower flow rate there is about a 1 CFM dif ference between the two flow gauges due to the filter-paper-induced pressure drop.

(2)- Safety Evaluation The safety evaluation included in 10 CFR 50.'59 safety analysis included as item f is also valid for this change. Likewise, the'goalis still the ~same: i.e.,

to have a- stack monitor sampling ' system .which makes it possible to determine if air leaks into the system have developed. Furthermore,' air leaks are stillindicated by observing changes in the a:r flow rates'shown on the -

intake and outflow air gauges. _ However, it is now clear that even when - 3

there are no air leaks into the stack monitor the two flow gauges ~do not read -

g Y

E 11 exactly the same. Equally important, however, is the fact that'when there-

~

- are no air leaks the difference between the readings on the two flow gauges -

is a known, constant value. Should this differet.ce increase, an air leak may be indiccted and can then be corrected if present.

A slight change in the internal diameter of the sampling probe has no effect on reactor safety. The velocity of the air in the probe and in the sampling line remains essentially the same and, therefore, sampling accuracy and the response time of the monitor remains the same.

The change in the air flow rate through the stack monitor necessitated a small change in the calibration factors for the monitor. Therefore, the probe change was followed immediately by the annual stack monitor calibration, which was about due anyway.

i

h. AUTOMATIC SHUT-OFF OF THE D102 HOOD FAN (1) Description A series of tests were performed which involved measuring air pressures in selected Radiatian Center and reactor building rooms with various building ventilation fans and hood fans on and off. - These tests showed that some _

degree of improvement in maintaining the reactor bay air pressure negative relative to surrounding rooms could be obtained if the D102 haod fan was turned off whenever the reactor bay ventilation fans went off. As a result =

of the test data, the reactor staff slightly modified the reactor building ventilation system to have the hood fan in D102 ' automatically shut' off 1 whenever the reactor bay velitilation fans go off. An alarm was installed in .

D102 to alert anyone using either hood or the pneumatic transfer system to -

the fact that trie hood fan has gone off. Personnel will be instructed in their-trairjng to exit D102 in such an event.

!4

-j u _ w

^

M g.1 , .

(2) Safety Evaluation

- There are no_ direct reactor _ safety implications associated with th's proposed '

change, and the change-does n affect the ventilation, system directly:

serving the reactor. However, the purpose of the change'was to increase overall radiation _ safety by ensuring that the air pressure in the reactor bay; remains negative relative to the air pressure in the adjacent rooms. The .

change helps to accomplish this objective which in tum helps to ensure that -

if there were a release of radioactive material into. the reactor bay there .

would be no leakage of the material out of tha bay.

q;  ;

2. 10 CFR 50.59 Chanaes to Reactor Procedures q

There were four changes to reactor procedures which were reviewed; approved, andI >

performed under the provisions of 10 CFR 50.59 during the reporting period.' i u

a, REVISIONS TO THE OSU RADIATION CENTER AND. TRIGA: ~ REACTOR $ J

.  :~

EMERGENCY RESPONSE PLAN AND EMERGENCY RESPONSE IMPLEMENTING -

PROCEDURES ,  ;

b (1) Description ;

~ ,

I A number of minor changes to the Radiation Center and.TRIGA Reactor-.

Emergency Response- Plan 1 and - Emergency 2 Response implementing :

p Procedures (ERIPs) were needed as a result of the annual exercise and the -

l annual review of the emergency plan. These changes are listed below?

Emeraency Plan Chanaes Title Page Tile last revision date was changed. from February to July 1991. -l li Page 3 2 . Under: Benton County Sheriff's Department, "the City.of . - >

LCorvallis Police Department" was ' replaced w'ith . ".OSU Police and Security."

b

=

?

L t r l j u - , _ . . . , ,-r. 4 - -, , - . - y ,. +

p .

. . 13-A 4

Under Benton County Emergency Services,1"Corvallis Fire =

or Police personnel" was replaced with "OSU Police and Security.";

Page 3-9 The list of. people potentially available for 'the rad'iological' assessment team was revise; to:

Geochemist ,

Health Physicist Neutron Activation Analysis Specialist OSU Radiation Safety Officer -

• Radiation Protection Technologist Radiation Specialists from the Radiation Safety Office

- Reactor Operator Reactor Supervisor ; -

Scientific Instrument Technician Senior Reactor Operater(s);

~

- Page 3-11 The line of succession for_ the Historian was revised to: ,

Radiation Center ' Office- Specialist E1: (Position 380),

Radiation Center Office: Specia!ist 1 (Position : 611),-

Radiation Center Office Specialist 2._

Page 3-14 - The Figure was replaced with a new one which has heavierc frame lines. -

/ ~

Page 7-8_ In 7.2.3 b), "OSU Campus' Security" _ was rep! aced with: .

"OSU.Pclice and Security" in two locations;

' Under Severe Natural Phenomenons a new paragraph 'was.

added: "'f the situation is severe and widespread, the OSU .

President may _ declare a disaster andlthelOSU Disaster -

-_ Management Plan will be put into offect; The OSU Disaster;

~

Management Plan is consistent with this Radiation Center :

and TRIGA Reactor Emergency Response Plan."-

\

,t i

f- ~ *

• rv f w

+

1

+ _, _ < .14e 3

-. c . <

g.. _4

~ Page 7-11 in 7 3.1-d), "OSU Campus Security"Lwas replaced-.with

- "OSU Police and Security." -

Page 81 in 8.1 d), "two telephones with seven outside lines," was -

replaced with "two independent telephones."

'f Page 8-3 in 8.2.2 b), " Radiation Center Stockroom" 'was replaced with " Health Physic's Laboratory."

~i in 8.2.3 a) . iii), " Room A138 of the Radiation -~ Center Complex" -was replaced with "the Health . Physics--

Laboratory."

f' Page 8-4 In 8.2.5 b), "at three positions" was deleted.

Page 8 5 In 8.2.5, d) was revised completely to_ read: ."A dual float operated microswitch is used to annunciate a low reac'.or.

tank water level. This is activated.when the water level

~

drops about two inchest- The"same floats activate a high water alarm when the reactor tank ' water level .ises about '

- one inch above normal."

Page'8 6- In 8.3.2, " Radiation Specialist" was replaced with."HealthI Physicist."  ;

. Page 8-8 8.4 a) i) was revised completely to read: ." Telephones:~ The ~

Radiation Center Complex 1 has approximately fifty i

-telephones, each with its own independent line. The main

^'

reception area (and hence ESC) telephones can handle eight [

simu!taneous calls;.and' include-~ displays fwhich provide?

information ~about the callsi' The telephones will coatinue? .

~

to function in the' event of a power loss to the Radiation--

Center." - .

3. ,

1 e .

~

, .c

. -15 =

Page 8 9 8.4 a) iii) was revised to read: " Portable Citizen's Band radios.

In 8.4 b), "OSU Campus Security" was changed to "OSU Police and Sect- y

• Page 10-1 In 10.1 b) v), "OSU Campus Security" was changed to "OSU Police and Security." -

Page A-3 Figure A.2 was replaced with a revised version which .

includes a grid system.

Page A-4 Figure A.3 was replaced with the latest revision of the RC floor plan.

Emeraency Resoonse imolementino Procedure Changg.g ERIA 3 (also OSTROP 1.)

Title Page Revision r: Umber was canged to 4 and the latest revision date was changed to 7/91.

Page IV.1.10 1.10 a was revised to read: "The reactor tank water level.

is monitored by dua: i: oats that are mechanically linked to; low water level and high water level microswitches. The.

~

floats will trip the appropriate microswitches if the water-level drops about two inches or rises about one inch."

Page IV.1.11 1.11 was revised completely to read:

T 1.11 Rod Withdrawal Prohibit

a. There arc two conditions which will result in rod withdrawal prohibit annunciation. -These are the low source count rate condition and the '

L i-

16-4 detector current selector switch (DCSS)'condi-tion as described below,

b. The count rate meter for the ~ reactor startup source is set to give the rod withdrawal prohibit annunciator when the count rate from the.

startup source detector decreases to about five counts per second.

c. When the DCSS is in a position to measure the current from the percent power channel or the safety channel using the picoammeter, the rod withdrawal orchibit annunciator is also lit. The '

swit#..s normally in one of these positions only during the performance of the reactor start up checks. Therefore, this condition is unlikely to occur during normal operation,

d. When the rod withdrawal prohibit annunciator-is energized, an interlock will prohibit control rod movement,

e. The console operator will not attempt to start the reactor with the rod withdrawal prohibit-annunciator energized,

f. If such an annunciator is received while the -

reactor is operating, the console operator will not be able to move the control rods, except to drop the rods in the process of scramming.

The console operator will:

1. Scram the reactor.

2. ' Notify the Reactor Supervisor.

\

1 q 17. =

7 (2) Safety Evaluation:-

Due to the fact that several of the textual changes previously' listed were the 3 result of a single common cause,it is appropriate to evaluate the previously.L listed changes by topic rather than individually.

Irs several places "OSU Campus Security" was replaced with "OSU Police and Security"'and in one location "the City of Corvallis Police Department" was replaced with "OSU- Police and Security." These changes were -

necessitated by the fact that the Oregon State Police recently assumed an -

armed response force role on campus. The changes in the emergency plan were consistent with those made in the physical security plan. They'merely -

update the plan and in no way reduce its effectiveness. - The number of" of ficers pote@lly allable and the police agency response time rernains the same.

The listing of those potentially available for the radiological assessment team was' revised to reflect changes in the job titles of the_affected personnel.

The quality of the people remains the same and their number is increased by one. Similarly, the line of succession for the Historian remains the same, but -

was updated to reflect the new state classification of their position titles.

These changes do ' reduce the effectiveness of the plan in any.way.

7 A number of the figures and personnellistings~ contained in the emergency.. -

plan and in the implementing procedures are in routine use at the Radiation; Center and are continually being updated.~ During the annual review of the t plan and procedures ~it is appropriate to add'the latest versions of these .

figures and listings. Two figures were changed as s' result of the~ 1991:

emergency exercise. - it was recognized that page 3-14 would be clearer to .

read if the frame lines were printed a little heavier than the other lines, and :

Figure A.2 was revised to add more detail and _ include grid lines.f These latter changes were made to_ make it easier to refer to particular locations 1 , ]

when making radiological measurements around= the RadiationiCenter Complex. - The grid lines arc not at uniform distances, but coincide with

1. SD

~

distinguishable features such as the edges of the eouds tand buildings. Tnese changes enhance the smergency plan-by enabling greater officiency in radiological assessmer't and radio communication.

The university now has a Disaster Management Plan and, therefore, this is now mentioned in the appropriate section of the emerge' icy plan. As noted, the Disaster Management Plan is consistent with the OSU Radiation Center and TRIGA Reactor Emergency Response Plan. Planning at the areater j university level can only enhance the university's response in the event of

]

an incident at the Radiation Center.

The Oregon ~ State System of Higher Eoucate completely changed its ,

e telephone system. This necessitated wording changed 5 the emergency plan. The new phone system is a significant improvement over the old system and will contribute positively to the effectiveness of the emergency response plan.

A rearrangement of offices and equipment resulted in extra radiological d assessment supplies being kept in the Health Physics Laboratory and not in I k

the stockroom. The amount and type of supplies remains the same, j Tho instrumented fuel element in the reactor has three thermocouples for i measuring fuel temperature. To operate the reactor, only one of these is '

j required by the Technical Specifications. Therefore, the reference to the -

three positions on page 8 4 was deleted. The required ability to measure-fuel temperature will always be maintair;cd, and hence the emergency plan effectiveness is not compromised by this change.

A new dual float system was installed to detect low and high reactor water '

levels. This necessitated wording changes to the relevant descriptive sections of the emercency plan,-but these _ changes do not reduce the effectiveness of the plan. The safety evaluation of the new float system was included in a previous 50.59 safety analysis, l

J

-19 The number of Citizen's Band radios is being increased and, therefore, the reference to a specific number was deleted. More radios will enhance communications capability and hence increase the effectiveness of the.

emergency plan.

The addition of the detector current selector switch (DCSS), necessitated a change to OSTROP 1 (ERIP 3). The DCSS addition was previously reviewed under a 50.59 safety analysis and, therefore, this was just a conforming procedural change.

b. CHANGE OF TIME INTERVAL BETWEEN TRANSIENT ROD CAllBRATIONS (1) Description Several years ago, the frequency of transient rod calibrations was changed from annual to semi-annual and then to quarterly. --The changes in the calibration frequency were rnade because the transient rod worth was increasing slightly as the core reactivity changed and the reactor was being routinely pulsed using the transient rod at or near the maximum allowable pulse reactivity insertion. Therefore, it was . felt that a more frequent' calibration was necessary to provide greater assurance that pulse reactivity insertions were accurate and- to ensure that the $2.55 limit on such reactivity insertions was not exceeded.

The reactor operations staff changed the transient' rod calibration frequency back'to annual.

(2) Safety Evaluation The original decision to increase the calibration frequency for the transient ' i rod was not safety related, but- was made out of a desire -to assure.

compliance with the Technical Specifications. Similarly, decreasing the calibration frequency for the transient rod is not related to safety, but was done because the potential for exceeding the pulse reactivity limit is no  !

I longer present.

l l

b.

. 20 First, the Table IV.D.1 shows that the transient rod worth has remained -

essentially constant within the measurement error of about i 20 for the past four years or more. Clearly, based on these data an annual calibration provides ten equally accurate assessment of current control rod worth.

Second, the reactor is no longer routinely pulsed to $2.55. With respect to this, there is a mechanical transient rod limiter set at $2.50, and the OSTR currently has an operational policy that pulses will be kept below about

$2.00. In addition, the reactor is now only pulsed a few times a year,

c. MINOR PROCEDURAL REVISIONS TO OSTROP 26, " PROCEDURES FOR THE USE OF EXTERNAL MONITORING AND RECORDING DEVICES" a

(1) Description The Scientific Instrument Technician purchased a small digital voltmeter for -  ;

use in the control room in association with OSTROP 26. This avoids the need for the reactor operations staff to borrow the technician's instrument.

However, in following OSTROP 26 as it was written,it was not possible to . I use the new voltmeter and obtain a stable zero reading during the initial te-t for computer voltage feedback to the reactor console. _The " phantom voltage" actually measured with the new voltmeter oscillated around zero due to the sensidve nature of the voltmeter and the high impedance at the 4

test location. As a result of this, OSTROP_26 was changed to require turning the TEST VOLTAGE switch to the ON position befoie testing the voltage. This action inserts a resistor in the test circuit and stabilizes the voltage, enabling it to be read accurately.

(2) Safety Evaluation l 1

The change has no unfavorable reactor safety implications, it was made to ensure that systems connected to-the reactor console for purposes of. i monitoring reactor parameters do not have any effect on the reactor. .The revision makes the test procedure more consistent and makes it'possible to read voltages more accurately because tu'r ning the TEST VOLTAGE switch q

l 1

21 to the ON position prior to testing voltages inserts a resistor in the test

, circuit, making the voltage more stable. The net effect of the proposed change increases safety.

d. TEMPORARY PROCEDURAL ADDITION TO OSTROP 26," PROCEDURES FOR THE -

USE OF EXTERNAL MONITORING AND RECORDING DEVICES" (1) Description As part of a reactor laboratory class (NE457) project, a variable operational amplifier was temporarily inserted between a fuel element thermocouple and no computer data acquisition system (CDAS). The reactor was then pulsed in an attempt to deterrnine any difference in the time delay of the peak reactor fuel temperature as measured by this system and as measured by the normal pulse temperature monitoring system. The amplifier increased the thermocouple signal from about 20 mV to about 7.5 V. Due to the fact that this activity was part of a class project, the addition to OSTROP 26 was temporary. The period of validity of the proposed change was 30 days from the date of approval as indicated by the signature of the ROC Chairman.

This time period has now passed.

(2) Safety Evaluation in normal operation of the CDAS during the pulsing _ mode, only one thermocouple is monitored. The modified procedure resulted in two thermocouples measuring the fuel temperature and, therefore, provided redundant fuel temperature data. The amplifier was only connected between the thermocouple and the CDAS and, therefore would not impact the reactor -

console or normal fuel temperature measurements during pulsing in any way.

In addition, all of the normal procedures requirud in OSTROP 26 were-completed, thereby providing further assurance that there was no influence:

of the CDAS on the reactor console.

4

D

.- 22 i 3. 10 CFR 50.59_Changas to Rostor Exneriments There were th oo changes to reactor experiments which were reviewed, approved, and performed under the provisions of 10 CFR 50.59 during the reporting period.

a. REVISION OF REACTOR EXPERIMENT B 23 (1) Descriptinn Reactor Experiment B 23 was revised in ordor to broaden the scope of the experiment. In particular, the revised experiment allows the insortion of shielding materials into any of the thermal columr. s removablo stringers to alter the gamma ray or neutron radiation fields. The experimont v'as also brought up to-dato by incorporating the applicable standard pt.ragraphs now used f or all reactor experiments, in addition, a grouttr variety of samplos are authorized for irradiation in the now experin int B-23. For examplo, small animals aru now specifically included, and there is no limitation on the mass of samplos inserted into thermal c.h>mn stringer locations Key provisions of the existing experlinent woro retained.

(2) Safety Evaluation The implications of incorporating the standard para 0raphs for experiments woro evaluated in previous modifications to experiments B 3 and B 11 and, thoroforo, are not ro ovaluated here. However, they cloarly enhanco safety by including porformance requirements common to all experiments.

The insertion of shieldin0 motorials to modif y the gamma and noutron fields, in general, reduces tho intensity of those fields and, hence, samplos irradiated in the thormbi column behind the shiolds receivo loss fluence than thoso irradiated under experiment B 3, Shielding materials, supporting equipment and samplos fully inserted into a thermal column stringer are still sufficiently for removed (27 inches of graphlto) from the robaor core to be noutronically decoupled from the core.

Thoretoro, thero are no measurablo reactivity of focts due to this experiment, l.

" ^" ^ - - ^ ~ ___w mm __

23-The activation of shielding materials, supporting equipment and samples irradiated in the thermal column will be estimated by the experimenter as part of the Irradiation Roquest. Thus, the reactor operations and health physics staf f s will be awaro of the potential dose ratos associated with these irradiated materials prior to performing an irradiation under this experiment.

Thorofore, theso individuals will be able to make appropriate arrangements  :

beforo removing the items from the thermal column. The health physics implications are, therefore, no dif forent than any other experiment which is run using the reactor.

Thoro is no need for a sample mass limitation in this experimont due to there being no reactivity considerations and due to the fact that the neutron fluxos aro low enough to virtually eliminate unexpected high levels of activation in irradiated materials. However, as noted previously, estimates of induced radioactivity in samplos, shleiding and supporting equipment will be included on the Irradiation Roc;uest which must be approved by the Senior Health Physicist and the Reactor Supervisor prior to perfotming the irradiation.

Irradiation of mall animals in a thormal column stringer has noicactor safety implications. All applicablo university regulations with respect to animal research will be folluwed,

b. REVISION OF REACTOR EXPERIMENT B 30 (1) Description Following a review of approved OSTR experiments, the Reactor Operations l Committee adopted the recommendations of the Reactor Administrator with respect to certain experiments. In addition to recommending that some -

experiments be moved to an inactivo status, there was a recornmondation that experiment B 30 be revised to incorporate applishle parts of the now standard wording relating to the conditions and' limitations imposed on reactor experiments. Experiment B-30 was, therefore, revised eccordingly, i

..m a. N..

24-The changes to Reactor Expenment B-30 are detailed below: i ta) Encapsulation in half dram vials was changed to 2/5 dram vials. 2/5 dram vials are the star dard vials used, and " half-dram" was probably just a colloquialism for these.

i (b) Information relating to the purpose of the irradiation (such as INAA)  :

was delt ted. The purpose of the irradiation is not rolovant to the safety issues. Raasons for irradiations have boon doloted from all of the newly revised and/or reformatted a experiments.

(c) The discussion of the original tost ir acht; Tis was deletod. Thee tosts were a condition of the original approvo:. They have been performed and results shown to be satisfactory many. years ago.

(d) Standard wording relevant to this experiment was included. The new standard wo< ding is present in the second, third and isst paragraphs in.

the now experiment description.

(2) Safety Evaluation The revised B 30 experiment is essentially the same as that which hr been approved, in exlstence, and in uso since 1976. Therefore, there are no additional safety issues requiring consideration. The incorporation of the.

. standard paragraphs slightly enhances safety by reminding experimenters of applicable Radiation Center and OSTR procedures and requirements.

c. SECOND 199192 REVISION OF REACTOR EXPERIMENT B 23 (1) Description The reactor staff was asked to amend Reactor Experiment B 23 to allow the contor irradiation cavity in the thermal column to be enlarged. When naoded, this onlargement can be achieved by removing the two graphite stringers directly adjacent to the thermal column's central stringer (stringer #5). One or both of the adjacent stringers can be removed allei the contral stringer

. 25- i (stringer #5) is removed. Once the central stringer is removed, it is very easy tr> slide out the two adjacent stringers. Therefore, the two adjacent stringers do not need to have threaded holes in their ends in order to remove them.

(2) Safety Evaluation ,

There are no undesirable reactor safety implications created by this change.

Because of the distance from the reactor core, any change in neutron reflection has a very minor offect on reactivity. Prior to the first use of the revised experiment the core excess reactivity was measured with no stringers removed and then with all three of the stringers removed to ,.

determine any reactivity effect due to the -hange. The difference was within the measurement error and therefore not significant, o

The thermal column shleid door is more than adequate to maintain low dose rates in the reactor bay even with the three stringers removed. _This was i demonstrated before the first use of the amended experiment by a radiation ,

survey in the area outside the closed thermal column door with the three stringers removed and the reactor at 1 MW.-

Dose rates at the outer face of the thermal column were higher with three stringers removed compared to only one removed; however, the same radiological controls are needed in both cases. Therefore, the change does not present a now situation, but researchers just wait a longer time after-reactor shutdown before retrieving experiments. Health physics staf f provide the access control until the radiation environment has been characterized and ,

a standard entry protocol estaboned.

1

-l N