ML20197B294

Annual Rept of Changes,Tests & Experiments for Jul 1985 - June 1986,per 10CFR50.59

ML20197B294
Person / Time
Site:	Oregon State University
Issue date:	10/03/1986
From:	Andrea Johnson Oregon State University, CORVALLIS, OR
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
NUDOCS 8610280345
Download: ML20197B294 (12)

Text

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9eenu U?r RFe". es University corvanis, oregon 97331 au .S Radiation Center < sos) 7s4-234 RI p q October 3,1%6cf/:j, U.S. Nuclear Regulatory Connission Region V 1450 Maria Lane, Suite 210 Walnut Creek, CA 94596-5368 Attention: Regional Administrator Gentlemen:

Subject:

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

The following report is submitted in accordance with the requirements of 10 CFR 50.59(b), and covers the OSTR's annual reporting period of July 1, 1985 through June 30, 1986.

During the specified reporting period there were three OSTR facility changes, two changes to the OSTR facility procedures, and one test conducted pursuant to 10 CFR 50.59. There were no changes to existing reactor experi-ments and no new experiments performed under the provisions of 10 CFR 50.59 during the current reporting period.

The individual changes and the one test being reported are listed below by category and by title, and are described in more detail in Attachment A. i Regarding this attachment, you will note that it includes a brief descrip-tion of each change or test, followed by a summary of the safety evaluation conducted for the described activity. As required, none of the changes or the test involved a change in the OSTR Technical Specifications or an unreviewed safety question as defined in 10 CFR 50.59(a)(2).

1. Changes to the OSTR Facility:

A. Temporary substitution for the normal reactor top continuous air monitor.

B. Removal of the non-freeze-stat for the steam-heated pre-heat coil in the reactor building air supply system.

C. Temporary installation of a radioactivity monitoring system -

in the OSU TRIGA reactor (OSTR) primary coolant loop.

2. Changes to OSTR Facility Procedures:

A. Power calibration procedural change.

B. Revision of the " General Limitations on Experiments Performed Using the OSU TRIGA Reactor" (as contained in OSTROP 18).

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October 3, 1986 USNRC 3. Tests Conducted Involving the OSTR:

A. Reactor power calibration test with a stirrer in the reactor tank.

We hope you find this year's report to be in good order. However, should you require more information or have any questions regarding our report, please let me know.

i erely, i d .. Ama i m v.. - -

Ar h Jr G. Johnson Act ng Reactor Administrator, OSTR Acting Director, Radiation Center AGJ/ef Enclosure cc: Document Control Desk, USNRC, Washington, D.C.

Director, Office of Inspection & Enforcement, USNRC, Washington, D.C.

Standardization and Special Projects Branch, Division of Licensing, USNRC, Washington, D.C., ATTN: Mr. Robert Carter Director, Oregon Department of Energy, Salem, OR B. Dodd, Assistant Reactor Administrator, OSTR S. E. Binney, Chairman, Reactor Operations Committee, OSTR T. V. Anderson, Reactor Supervisor, OSTR STATE OF OREGON )

)ss COUNTY OF BENTON )

A. G. Johnson, being first duly sworn on oath, deposes and says that he has affixed his signature to the letter above in his official capacity as Acting Reactor Administrator; that in accordance with the provisions of Part 50, Chapter 1, Title 10 of the Code of Federal Regulations, he is attaching this affidavit; that the facts set forth in the within letter and attachment are true to his best info ation and belief.

u @6 A. G. IJo inson Acting @actorAdministrator Subscribed and sworn to before me, a Notary Publie n in and for the County of Benton, State of Oregon, this 7'M day of /(ATM ,

A.D., 1986.

/l btd SI2&lf0 Notdry Public oV0regon My Commission Expires

x ATTAC}9ENT A CHANGES TO THE FACILITY, TO FACILITY PROCEDURES, TO REACTOR EXPERIDENTS, AND TESTS CONDUCTED PURSUANT TO 10 CFR 50.59 FOR THE PERIOD JULY 1, 1985 THROUGH JUNE 30, 1986

1. Introduction The information contained in this attachment provides a summary of the changes and tests performed during the reporting period under the provisions of 10 CFR 50.59. As applicable, the items to be reported have been grouped into three categories: those dealing with changes to the facility itself; those dealing with changes to the facility's procedures; and those involving tests and changes to OSTR experiments.

For each item identified, a brief description of the action taken (i.e., a change or test) and a summary of the safety evaluation are included.

2. 10 CFR 50.59 Changes to the Facility There were three changes to the facility itself which were reviewed and performed under the provisions of 10 CFR 50.59 during the reporting period. A summary of each change and its safety evalu-ation follows,

a. Temporary Substitution for the Normal Reactor Top Continuous Air Monitor Description The dual channel (gas and particulate) reactor top continuous air monitor (CAM), which is normally used at the OSTR for reactor-top air monitoring, was needed for a graduate student research project directly related to 41 rArelease from the OSTR tank. Therefore, it was removed from operational service and replaced with a single (particulate) channel CAM.

The OSTR Technical Specifications require only a particulate CAM to be operational on the reactor top. The substitute single channel CAM had been previously used at the OSTR for many years to moni. tor the air on the reactor top prior to the purchase and installation of the newer dual channel CAM.

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Safety Evaluation ,

The described air monitor change restores the reactor top air monitoring system to a former mode of operation and does not involve any safety implications or constitute an unreviewed safety question. The gaseous activity evolving off the reactor water has been well characterized in recent years and is known not to present any significant hazards in approved operating modes.

b. Removal of the Non-Freeze-Stat for the Steam-Heated Pre-Heat Coil in the Reactor Building Air Supply System Description The steam-heated pre-heat coil for intake air in the reactor building air supply system has been equipped since initial instal-lation with a non-freeze-stat designed to prevent freezing of any steam condensate present in the pre-heat coil. This device, which is briefly described on page 2-13 of the 1968 OSTR Safety An'alysis Report, has been ineffective in preventing condensate freezing and was removed by OSU Physical Plant. The removal of this non-freeze-stat was carried out as part of a heating system remedy designed to eliminate condensate freezing in the pre-heat coil by providing more steam to the coil on a continuous basis during periods when incoming air temperatures are unusually low.

Safety Evaluation The non-freeze-stat has never had any impact on reactor building heating or ventilation. Thus, its removal will cause no changes whatsoever, and there are absolutely no safety implications or unreviewed safety questions associated with the freeze-stat removal and the related efforts to get more steam in the pre-heat coil.

The reason for completing a 10 CFR 50.59 safety evaluation is based on the fact that the non-freeze-stat was briefly described in the currently applicable OSTR Safety Analysis Report and, therefore, its removal constitutes a minor facility change.

c. Temporary Installation of a Radioactivity Monitoring System in the OSU TRIGA Reactor (OSTR) Primary Coolant Loop Description A microcomputer-controlled sampling system was developed to peep 25 m1 to 100 ml TRIGA primary coolant water samples into a standard reference geometry surrounding a Ge(Li) detector. After radionuclide analysis of the water sample, the system design made it possible to route the sample to a liquid hold-up container or to a liquid waste container.

Another part of the system was configured as an on-line monitor using a NaI(T1) detector to measure the gross gamma count rate from the water in the primary coolant loop. Based on this on-line monitoring feature, when the primary water gross gamma count rate exceeded a user-selected threshold, a water sample was automatically tapped from the primary coolant loop for analysis with the Ge(Li) detector system.

Installation of the system in the OSTR primary coolant loop was designed so as to use the upper annubar flow element in the heat exchanger room. Flows through the system were limited to 40 ml/ min by adjustment '

of the flow meter tap valve. The tap was closed at the end of each work session. The total volume of coolant sampled during these experiments did not exceed 10 liters. The system was equipped with an appropriate liquid waste container to hold the coolant after analysis and before disposal.

All connections in the sampling system were threaded fittings, the systems rested in a plexiglass pan with 15 liters of capacity, and a catch pan was placed below the valve tap to collect any drips.

The sampling system and a small shielded Ge(Li) detector were placed on a wheeled table adjacent to the west wall of the reactor bay, which was also near the coolant sample tap. A rack containing the microcomputer was placed nearby. A 40 psi air line for operation of the system's valves was run from a reactor bay compressed air manifold to the sampling system /Ge(Li) detector cart on the west wall.

The system remained installed for several weeks to allow time to sample coolant during various patterns of reactor operating history and to demonstrate the automation capabilities.

4 Safety Evaluation Incorporation of this system did not have any undesirable impact upon the safety features of the OSTR, and there were no unreviewed safety questions involved. On the other hand, additional information regarding radioactivity in the primary coolant was viewed as a positive safety benefit of this effort. The normal OSTR primary water radioactivity monitor remained in use during this period.

3. 10 CFR 50.59 Changes to Facility Procedures There were two changes to facility procedures reviewed and ap-proved under 10 CFR 50.59 during the reporting period. A description of these changes follows.

b. Power Calibration Procedural Change Description The OSTR power calibration procedure was changed as a result of the test described under section 4.a. of this attachment, "10 CFR 50.59 Tests and Changes to Reactor Experiments," which follows this section. In particular, the changes involved the tank factor, which was changed from 0.0525 C/kWh to 0.0493 C/kWh, and an additional requirement that the power , calibration be performed with a stirrer in the reactor tank.

Safety Evaluation The mechanical aspects of using a stirrer in the reactor tank have been addressed in the 50.59 safety analysis dealing with the power calibration test (see section 4.a.). The combined effect of using the new tank factor and the stirrer is that the previous reactor power was assessed as being about 3% low. Therefore, calibration using the new procedure has no safety implications and involves no unreviewed safety questions, and yet provides a more accurate indication of the actual reactor power.

c. Revision of the " General Limitations on Experiments Performed Using the OSU TRIGA Reactor" (as contained in OSTROP 18)

Description Oregon State TRIGA Reactor Operating Procedure (0 STROP) 18 deals with the process of approving and performing experiments involving use of the OSTR. Included within OSTROP 18 is a section entitled " General Limitations on Experiments Performed Using the OSU TRIGA Reactor." The information in this specific section is also printed as a separate document with the same title and is distributed to users and to potential users of the reactor.

Two changes to this document were made during the current reporting period. The first revision involved mainly editorial changes which clarified the limitations while keeping them essentially unchanged.

The second revision to the " general limitations" information focused mainly on item 17, which deals with the encapsulation requirements for samples which are submitted for irradiation in the OSTR. However, in addition to item 17 revisions, a change to item 18 was also included. These modifications are explained below.

1) Description of Revisions to Item 17:

Following several recent reviews cf the OSTR sample encapsulation requirements by the reactor operations staff, the staff concluded that the requirements needed certain changes. In particular, it appeared that the encapsulation requirements would be improved by several editorial additions to clarify their intent and by adding a second specific option for double encapculation of samples submitted for irradiation in excess of 1 MWhr. The editorial changes to the containment requirements were mostly minor and involved no changes to the presently approsed encapsulation policy.

However, they now specifically state the need for advance approval of certain encapsulation techniques by the Reactor Supervisor and the Senior Health Physicist, and they clarify the policy for encapsulation of samples during " soaker" irradiations (irradiations longer than 13 MWhrs). Since the editorial additions simply reflect what has been actual practice in the OSTR operations program for

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nearly 19 years, the addition of this information introduces no changes, but the presence of such information helps to alert experimenters to our requirements, and therefore enhances safety by increasing the likelihood that proper encapsulation will be achieved.

Limiting the approved method of double encapsulation for sample irradiations greater than 1 MWhr through 13 MWhrs to the currently approved double-heat-sealed polyvial technique has created unnecessary difficulties for researchers who need to irradiate samples occupying a volume greater than about 1 ml. As a result, the reactor operations staff investigated alternate double-encapsulation techniques which would allow larger sample volumes to be irradiated while still maintaining adequate double containment. The staff investigation ultimately involved actual tests of one new double-encapsulation process (see Safety Evaluation section), which used a standard heat-sealed two-dram polyvial for the inner container (in order to allow a larger sample volume) and a heat-sealed plastic bag for the outer container.

The two-dram polyvial was selected as the inner container in this case because it is the next larger size vial above the largest vial currently approved as an inner container, and this particular size vial has shown excellent containment characteristics as an outer container in the currently approved double-encapsulation procedure for greater than 1 MWhr irradiations. However, the staff sees no reason why any of the heat-sealed polyvial sizes of two drams or smaller cannot be used as the inner sample container inside an approved heat-sealed plastic bag, since all sizes below two drams are already approved as inner-sample containers. (However, note that plastic bags with inner vial containers smaller than two drams are not likely to be used extensively since the smaller polyvials can be easily fit into a two-dram vial, which is easier to work with and is now our standard method of double encapsulation.)

Heat-s> slable plastic bags were selected as the optional outer container since there are no readily available polyvial si7.es large enough to easily contain the two-dram polyvial chosen l

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for the inner container. However, as a result of staff tests, the only heat-sealable plastic bag found acceptable, based on containment integrity after 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> at 1 MW in the rotating rack, was the Nalgene liquid scintillation bag in the 10 m1 size. In spite of the current tests, other heat-sealable bags may be available which will prove satisfactory after testing, but for the present the staff recommends specifying the 10 ml Nalgene bag as currently acceptable and allowing use of other such bags if they successfully demonstrate their ability to contain samples after at least 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> at.1 MW in the reactor's rotating rack irradiation facility.

Records of all such container tests have been and will continue to be kept by the Reactor Supervisor, and approval of results must be jointly obtained from the Reactor Supervisor and the Senior Health Physicist before the containment is acceptable for use.

As a matter of conformity, and based on this 10 CFR 50.59 evaluation, Irradiation Requests (irs) were modified to add a small space for experimenters to briefly designate the type or types of sample encapsulation used. The presence of this information will then allow the Reactor Supervisor and the Senior Health. Physicist to review the encapsulation method prior to sample irradiation and to document their required approval of the encapsulation process used, along with other required approvals, by signing the IR.

2) Description of Revision to Item 18:

Item 18 addresses the approval requirements for irradiation of materials which might introduce conditions requiring special safety precautions, such as higher-than-normal radiation levels.

Current limitations on the irradiation of such materials are already addressed in various specific reactor experiments, and item 18 itself now specifically requires prior approval of the Reactor Supervisor, the Senior Health Physicist, and the Assistant Reactor Administrator. While it is clear that the opinion of these three reactor operations staff members will weigh heavily on any decisions to irradiate or not irradiate such materials, the reactor staff believes that item 18 should add the requirement that a quorum of the Reactor Operations Committee provide final approval for these irradiations, and thus such a change was added.

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Safety Evaluation

1) Revisions to Item 17:

The editorial revisions to the general limitations in no way changed current encapsulation policy, which has been used successfully in the OSTR operation for nearly 19 years. The changes, in fact, contribute further to a safe operation by making the encapsulation requirements clearer to reactor users and by clarifying the approval process for various encapsulation options.

The proposed adoption of a new double encapsulation option for irradiations exceeding 1 MWhr up through 13 MWhrs was suggested based on a long and successful experience with the containment characteristics of the heat-sealed two-dram polyvial and upon actual tests of the Nalgene liquid scintillation bags.

The bag tests were conducted with six heat-sealed two-dram polyvials filled one-half to three-fourths full of water inside six individual heat-sealed Nalgene liquid scintillation bags.

This double-encapsulation method was then irradiated for 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> at 1 MW in the OSTR rotating rack. The results showed no_ leaks in any of the polyvials and no leaks in any of the plastic bags.

The bags did not discolor or become brittle, and appeared to be quite able to withstand further irradiation.

In view of the fact that the two-dram vials rarely leak after 13 MWhrs of irradiation, and based upon the integrity of the heat-sealed Nalgene bags, the staff saw no risks involved in adopting this new double-encapsulation technique as an added option, and further concluded that none of the above changes involved an unreviewed safety question.

2) Revision to Item 18:

The revision to item 18 simply increased the level of approval which must be obtained in order to irradiate certain materials.

This can only help improve safety by requiring a slightly more formal review of any needed safety precautions, and therefore the change involves no unreviewed safety questions.

4. 10 CFR 50.59 Tests and Changes to Reactor Experiments One test was reviewed, approved and performed under 10 CFR 50.59, but there were no new experiments and no 10 CFR 50.59 changes to existing reactor experiments during this reporting period. The test involved is described below.

a. Reactor Power Calibration Test with a Stirrer in the Reactor Tank Description In order to support possible changes to the OSTR power calibration procedure, it was necessary to carry out a pre-planned test involving an actual power calibration. The test was performed with a mechanical water stirrer in the reactor tank and with the. reactor primary water circulation system turned off.

The test itself involved a square wave power increase to 1 MW followed by temperature measurements at one minute intervals for ten minutes, after which time the reactor was scrammed. Before, during, and after the 1 MW operation the stirrer was used to mix the water in the reactor tank. The stirrer was inserted on the west side of the reactor tank and was suspended from the crane by a 3-ton test nylon sling. It was then clamped to the reactor tank cover to prevent rotation and to provide additional support.

Another rope was also attached to the stirrer as an extra safety line to prevent its dropping into the tank.

Safety Evaluation The reactor was operated in a normal mode and even though the primary water circulation system was off, the short duration of the test ensured that specified fuel temperatures and primary water temperatures remained well within limits. The need for the 10 CFR 50.59 analysis exists because of the use of the stirrer and the use of the crane over the tank during reactor operation.

The stirrer's propeller was well clear of any reactor components.

As noted previously, the stirrer was firmly supported by the crane and the clamp, and also had an additional safety line. Therefore, the probability of dropping the stirrer into the tank was extremely l small. In addition, such stirrers are conventionally used by non-power reactors when performing required power calibrations.

No access to the reactor bay was allowed during the duration of the test to prevent unnecessary personnel exposure from increased radiation levels due to 16N and 41A r. These increased levels (due mostly to 16N) arose because the diffuser system, which delays the rise of 16N in the OSTR tank, goes off when the primary water circulation system is turned off.

As a result of the above considerations, it was concluded l that there were no unreviewed safety questions associated with this test and that it .41d be safety performed under the provisions of 10 CFR 50.59.

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