ML11347A363
| ML11347A363 | |
| Person / Time | |
|---|---|
| Site: | Texas A&M University |
| Issue date: | 12/02/2011 |
| From: | Juzaitis R Texas A&M Univ |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC ME1588 | |
| Download: ML11347A363 (47) | |
Text
DWIGHT LOOK COLLEGE OF ENGINEERING Department of Nuclear Engineering AIM ý TEXAS A&M U N I V E R S I T Y December 2, 2011 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Docket No.50-059
SUBJECT:
Response to "Texas A&M University - Request for additional information regarding the Texas A&M University AGN-201M reactor license renewal application (TAC. NO. ME1588), July 25, 2011" In response to the RAI dated July 25, 2011 Texas A&M University is submitting answers to the following questions: 7, 12, 18, 19, 20, 21, 22, 23, 24, 25, 27, 28, 29, 30, 31, 33, 34, 35 and 36.
The application for License Amendment No. 10 is also included as supplemental information.
The remaining questions will be submitted prior to December 31, 2011.
If you have any questions, please do not hesitate to contact me at: (979) 862-1956, or e-mail at rjuzaitis(atamu.edu.
I declare under penalty of perjury that the foregoing is true and correct. Executed on December 2, 2011.
337 Zachry Engineering Center 3133 TAMU College Station, TX 77843-3133 Tel. 979.845.4161 Fax. 979.845.6443 nuclear.tamu.edu
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- 7. NUREG-1537, Part 1, Section 7.2.2, "Design-Basis Requirements," requests the function or purpose of systems or instruments that monitor reactor parameters. SAR Sections 7.1 and 7.2.3 state that the skirt monitor is intended to scram the reactor and sound an evacuation alarm. Please include a limited condition for operation (LCO) and surveillance requirement for the skirt monitor scram and alarm or justify why this scram and evacuation alarm are not required to be included in Technical Specifications (TS). Please describe how often the skirt monitor detector is calibrated and how the setpoint is derived and implemented.
Proposed LCO 3.4.g The skirt monitor interlock shall be set to prevent reactor startup and scram the reactor if skirt radiation levels exceed 2 times the last recorded 5 watt value.
Proposed Surveillance requirement Modify T.S. 4.4.b 4.4.b Prior to each day's reactor operation or prior to each reactor operation extending more than one day, the reactor room high radiation area alarm and the skirt monitor interlock shall be verified to be operable.
Add T.S. 4.4.d 4.4.d The skirt monitor interlock shall be calibrated and set annually.
Calibration The skirt monitor is calibrated annually. The setpoint is twice the 5 watt reading, which is set on the skirt monitor alarm module with the use of an independent electrical signal.
- 12. NUREG-1537, Part 1, Section 7.2.2, "Design-Basis Requirements," requests a description of the function or purpose of systems or instruments. SAR Section 7.5 describes a criticality monitor for the source locker. Please provide a description of this monitor and its purpose and indicate if this monitor is required by the reactor license.
This monitor is not required to be installed at the facility pursuant to 10 CFR 70.24(a). Our facility license limits our special nuclear material inventory below the levels set forth in 10 CFR 70.24(a).
The installed monitor is a Nucleus Model L transistorized count rate meter. The meter monitors radiation levels inside the source locker. The meter has full range scales of 500; 2,000; 5,000; 20,000; and 50,000 counts per minute. An audible and visual alarm is also provided for the meter.
- 18. NUREG-1537, Part 1, Section 12.1.2, "Responsibility," requests discussion of responsibilities for the safe operation of the reactor and the reactor facility for individuals that appear in the organization structure. SAR Section 12.1.1 description of the Reactor Supervisor responsibilities does not include the SAR Section 10.3 responsibility that experiments will be conducted under the direct supervision of the Reactor Supervisor. Please verify that this responsibility is one of the Reactor Supervisor responsibilities, if this responsibility is permitted to be delegated and if delegated by what mechanism.
The Reactor Supervisor is responsible for the conduct of experiments. This responsibility is permitted to be delegated to the senior reactor operator of record.
Modify T.S. 6.1.4 to include:
"The Reactor Supervisor or the designated senior reactor operator of record shall directly supervise experimental procedures."
- 19. NUREG-1537, Part 1, Section 12.2.2, "Charter and Rules," requests discussion of requirements for a quorum when voting. SAR Section 12.2.2 states the operating organization will not comprise a voting majority of the members at any Reactor Safety Board. Please discuss how this voting control is met, specify by title which persons are considered members of the operating organization and propose changes to TS 6.4.1 that incorporate these details or provide an explanation describing your reasons for not incorporating the changes.
The operating organization is comprised of the Reactor Supervisor and any licensed reactor operator(s). This condition is met by having the Reactor Supervisor serving only as an ex officio member of the Reactor Safety Board with no voting ability. In the event that any voting member(s) of the Reactor Safety Board were part of the operating organization, the chairman would be responsible to ensure that a majority of those members voting were not part of the operating organization.
Modify T.S. 6.4.1 The Reactor Safety Board shall meet as often as deemed necessary by the Reactor Safety Board Chairman but shall meet at least once each calendar year. A quorum for the conduct of official business shall be the chairman, or his designated alternate, and two (2) other regular members.
The chairman, or his designated alternate, shall ensure that the operating organization does not comprise a voting majority of the members at any Reactor Safety Board meeting. The operating organization is comprised of the Reactor Supervisor and any licensed reactor operator.
- 20. NUREG-1537, Part 1, Section 13, "Accident Analyses," states that the NRC staff has generally found acceptable that doses to facility staff for accident analysis results are less than 5 rem whole-body. SAR Section 13.1.1 provides a Maximum Hypothetical Accident (MHA) with calculated results that exceed the acceptable limit for the occupationally exposed staff member. Please provide an evaluation of a safety analysis of the MHA for your facility that indicates the resulting occupational doses are within the Total Effective Dose Equivalent limits of 10 CFR 20.1201 for an occupationally exposed staff member.
Please provide an evaluation of a safety analysis, with calculations, for:
- a. The maximally exposed staff member.
- b. The staff member at the reactor console, and;
- c. The potential direct dose to a person in the unrestricted area closest to the reactor room and show the dose is within the limits of 10 CFR 20.1302, compliance with dose limits for individual members of the public.
License Amendment No. 10 outlined the MHA to result from a 1% step change in reactivity.
This scenario was accepted in the SER for Amendment No. 10. The peak power from this excursion would be 2.66 x 103 watts, with total energy release of 74.59 watt sec. The evaluation also states that the highest dose to members of the general public in the unrestricted area would be 0.0366 mRem located in an adjacent laboratory. This data was used to evaluate the maximally exposed staff member and the staff member at the reactor console.
Using the know radiation levels at 0, 1, 3, and 5 watts, a linear equation was derived. This equation was used to extrapolate data for the accident scenario. The following assumptions were made: evacuation time of 3 minutes and a constant radiation field. The maximum exposed staff member would receive about 2.25 Rem TEDE. This represents a staff member located at the point 7 on the west edge of the reactor. The operator would receive approximately 22 mRem TEDE.
- 21. ANSJIANS-15., Section 1.3, "Definitions," provides definitions commonly used in Research and Test Reactors. TS 1.0 contains several definitions that do not conform or lacked recommended detail. Please propose changes to the definitions that conform to ANSI/ANS-15.1 guidance for: reactor secured, reactor shutdown and shutdown margin (SDM) or provide an explanation describing your reasons for not incorporating the changes.
As of the most recent revision, TS 1.0 was updated with the definitions meeting the requirements of ANSI/ANS-15, Section 1.3 with one exception. The following change is proposed to meet this exception:
1.26 Shutdown Margin - Shutdown margin shall mean the minimum shutdown reactivity necessary to provide confidence that the reactor can be made subcritical by means of the control and safety systems starting from any permissible operating condition with the most reactive rod and the fine control rod in its most reactive condition, and that the reactor will remain subcritical without further operator action.
- 22. NUREG-1537, Part 1, Appendix 14.1, Section 3.2(4), "Scram Channels," states that historically, there have been cases in which the NRC has accepted power level scrams higher than licensed power (1.2 times licensed power level is common) if supported by the safety analysis. TS 3.2.e, Table 3.1 lists the high power scram setpoint as <10 watts. Please consider revising the high power scram setpoints in TS Table 3.1 to 120% of licensed power (6 watts) or justify by safety analysis why this scram setpoint should remain as proposed (10 watts).
T.S. Table 3.1 will be modified as follows:
Nuclear Safety #2 (Log Power)
High Power
< 6 watts scram at power > 6 watts Nuclear Safety #3 (Linear Power)
High Power 6 watts scram at power >6 watts
- 23. ANSI/ANS-15.1, Section 1.3, "Definitions" requests the (SDM) be made with the nonscrammable rods in the most reactive position. In TS 3.1.b, the (SDM) determination does not include the reactivity of the fine control rod, a non-scrammable rod, in the most reactive position (inserted). Please propose changes to TS 3.1.b to include the reactivity effect of the fine rod failing in its most reactive state to the SDM determination or provide an explanation describing your reasons for not incorporating the changes.
Modify T.S. 3.1.b 3.1.b The shutdown margin with the most reactive safety or control rod and the fine rod fully inserted shall be at least 1% A k/k.
- 24. NUREG-1537, Part 1, Appendix 14.1, Section 3.2(4), "Scram Channels," requests a list of all required scram channels. TS 3.2 and TS Table 3.1 do not include the Interlock Relay Scram in TS that is described in the SAR Section 7.3. Please propose changes to TS 3.2 and Table 3.1 adding the Interlock Relay Scram as an LCO with corresponding surveillance requirement or provide an explanation describing your reasons for not incorporating the changes.
The Interlock Relay Scram is a portion of the safety system that includes the following scram functions: Shield Water Temperature switch, Earthquake switch, Shield Water Level switch, Channel 1 Low Count Rate, Channel 2 Low Level, Channel 3 Low Level, Rod Drive System Plug and Relay Chassis Interlock. Each scram is discussed below.
Shield Water Temperature switch - This interlock is addressed in T.S. 3.2.g.
Earthquake switch - This interlock is addressed in T.S. 3.2.h.
Shield Water Level switch - This interlock is addressed in T.S. 3.2.f.
Channel 1 Low Count Rate - This interlock is addressed in T.S. 3.2.d.
Channel 2 Low Level - This interlock is addressed in 3.2.d.
Channel 3 Low Level - This interlock is addressed in 3.2.d.
Rod Drive System Plug - This interlock was not incorporated into the new control console. With the new console design, if an individual rod drive plug is disconnected, indication for that rod is completely lost. In the event this occurs, the operator will perform a manual scram of the reactor. The facility operating procedures will be modified to reflect this action.
Relay Chassis Interlock - This interlock was not incorporated into the new control console. With the new console design, if the input signal to any instrument is lost and the reactor does not scram on this loss of signal, the operator will perform a manual scram of the reactor. The facility operating procedures will be modified to reflect this action.
- 25. ANSI/ANS-15.1, Section 4.2(5), "Reactor Control and Safety Systems," requests appropriate surveillance testing (e.g., operability checks, calibrations, and inspections) for scram channels. TS 3.2.g states that a seismic displacement interlock switch shall be installed in such a manner to prevent reactor startup and scram the reactor during a seismic displacement.
Please propose and justify a setpoint for this switch and propose changes to TS to provide a surveillance for this switch or provide an explanation describing your reasons for not incorporating the changes.
The seismic displacement interlock is designed to cause a reactor shutdown in the event the instrument receives a horizontal acceleration that causes a horizontal displacement of 1/16 inch or greater. The facility currently has no method to quantify the magnitude of displacement for this device. The sensor has proven to be quite sensitive, activating when the reactor is simply struck with a rubber mallet. This interlock is tested semiannually currently. This test has been deemed adequate to ensure that safety is provided for during a seismic event.
- 27. ANSIANS-15.1, Section 3.8.3, "Failure and malfunctions," requests that failure of any experiment shall not result in releases or exposures in excess of 10 CFR 20 limits. TS 3.3.d.
includes a two hour exposure for persons in the unrestricted area starting with the time of a release. Please provide an evaluation of a safety analysis showing persons can be removed from the exposure in the specified time and that the specified time is consistent with evacuation times in your Emergency Plan.
Question removed via phonecom with license renewal PM, Walt Meyer.
- 28. ANSIIANS-15.1, Sections 4.2(1) and 4.2(4), "Reactor Control and Safety Systems," request measurement of scram time and rod reactivity worth be performed after any work is done on the rods or the rod drive systems. TS 4.2.a. does not include the requested measurement of scram time and rod reactivity worth after work is done on reactor rods and rod drives. Please propose changes to TS adding these requirements to TS 4.2.a or provide an explanation describing your reasons for not incorporating the changes.
Modify T.S. 4.2.a 4.2.a Safety and control rod scram times and average reactivity insertion rates shall be measured annually and following work on the rods or the rod drive systems, but at intervals not to exceed 15 months.
- 29. ANSI/ANS-15.1, Section 3.1 (4), "Core configurations," requests LCO specifications be included for core configuration components that assure the assumptions for accident analyses are maintained. The Maximum Hypothetical Accident scenario assumes that no fission products escape the core tank. Please propose changes to TS 3.4.f to include the specification that "the core tank shall be sealed during reactor operations" or provide an explanation describing your reasons for not incorporating the changes.
There are currently no plans in place to perform maintenance or alter the core tank, thus the core tank remains sealed at all times. If any maintenance or inspection does occur on the core tank, the core tank shall be sealed prior to the conclusion of the procedure and prior to the next reactor operation.
Including the specification "the core tank shall be sealed during reactor operations" to TS 3.4.f results in an issue verifying the requirement is fulfilled during reactor operations. TS 3.4.f. 1 is confirmed by a water level float and associated low level trip that continuously ensures the requirement is fulfilled. TS 3.4.f.2 is verified prior to each day's reactor operation by procedure 3.3.6 of the Standard Operation Procedures. However, there is no system in place that allows regular verification of a requirement specifying the core tank shall be sealed during reactor operations.
- 30. ANSI!ANS-1 5.1, Section 4.0, "Surveillance requirements," requests appropriate surveillance testing (e.g., inspections) for reactor component operability. TS 4.3.b. specifies that a visual inspection for water leakage from the shield water tank shall be performed annually and leakage shall be corrected prior to subsequent reactor operation. Please provide an evaluation of the proposed inspection frequency and propose changes to TS 4.3.b to perform a visual inspection or check more frequently or provide an explanation describing your reasons for not incorporating the change.
Prior to the first reactor operations of the day, the shield tank will be inspected for leakage. This frequency along with the shield tank low level alarm will ensure that reactor operations do not commence without proper shielding from the shield tank. If any leakage is observed, it will be corrected prior to reactor operations. This would include but not be limited to; cleaning up any leakage, investigating the reason for leakage, correcting the deficiency, and filling the tank if necessary.
Modify T.S. 4.3.b Visual inspection for water leakage from the shield tank shall be performed prior to the first startup of the day. Leakage shall be corrected prior to subsequent reactor operation.
- 31. ANSI!ANS-15.1, Section 5.1, "Site and facility description," requests a description of the site and reactor facility. TS 5.1 describes the Reactor Room and Accelerator Room but does not specify the reactor licensed area. Please define in TS 5.1 the areas, including room numbers where appropriate, that are under the reactor license.
Modify T.S. 5.L.a 5.1.a The reactor room, located on the ground floor of the Zachry Engineering building, room 61, houses the reactor assembly, accessories required for its operation and maintenance, and the reactor control console.
Modify T.S. 5.L.b 5.1.b The accelerator room, located on the first floor of the Zachry Engineering building, room 135, is directly above the reactor room and a hole in the accelerator room floor provides access to the thermal column.
- 33. ANSI/ANS-15.1, Section 6.1.3(3) "Staffing," describes events when the presence of an SRO is required. TS 6.1.9 does not include 1) initial startup and approach to power after the reactor has been secured and 2) recovery from unplanned or unscheduled shutdown or significant power reduction. Please include these events in TS 6.1.9 as these are required by 10 CFR 50.54(m)(1).
Modify T.S. 6.1.9 6.1.9 Operating Staff
- a. The minimum operating staff during any time in which the reactor is not shutdown shall consist of:
- 1. One licensed Reactor Operator at the reactor control console.
- 2. One other person in the reactor room certified by the Reactor Supervisor as qualified to activate a manual scram and initiate emergency procedures.
- 3. One licensed Senior Reactor Operator readily available on call. This requirement can be satisfied by having a licensed Senior Reactor Operator perform the duties stated in paragraph 1 or 2 above or by designating a licensed Senior Reactor Operator who can be readily contacted by telephone and who can arrive at the reactor facility within 30 minutes.
- b.
A licensed Senior Reactor Operator shall supervise initial startup and approach to power.
- c.
A licensed Senior Reactor Operator shall supervise recovery from unplanned or unscheduled shutdown.
- d.
A licensed Senior Reactor Operator shall supervise all reactor maintenance or modification which could affect the reactivity of the reactor.
- e.
A listing of reactor facility personnel by name and phone number shall be conspicuously posted in the reactor control console area.
- 34. ANSI 15.1, Section 6.2.2. "Charter and rules," requests a timely dissemination, review and approval of minutes. TS 6.4.5, does not specify to whom the Reactor Safety Board minutes are distributed or a timeframe for the distribution. Please propose changes to TS 6.4.5. that provide the requested information or provide an explanation describing your reasons for not incorporating the changes.
Modify T.S. 6.4.5 6.4.5 Minutes of the Reactor Safety Board The Chairman of the Reactor Safety Board shall direct the preparation, maintenance, and distribution of minutes of its activities. These minutes shall include a summary of all meetings, actions taken, audits, and reviews. Distribution of the minutes shall be to all Reactor Safety Board members and ex-officio members within 60 days of adjournment.
- 35. ANSFANS-15.1, Section 6.6.2, "Actions to be taken in the event of an occurrence of the type identified in Sections 6.7.2(l)(b) and 6.7.2(1)(c)," describe actions to be taken for reportable occurrences. TS 6.9.2 has no specified actions required for reportable occurrences. Please propose changes to TS 6.9.2 to include the actions specified above for reportable occurrences or provide an explanation describing your reasons for not incorporating the changes.
Modify T.S. 6.9.2 as follows:
6.9.2 Reportable Occurrences In the event of an occurrence outlined below, the following actions shall take place.
Reactor conditions shall be returned to normal, or the reactor shall be shut down. If it is necessary to shut down the reactor to correct the occurrence, operations shall not be resumed unless authorized by the Head of the Department of Nuclear Engineering or designated alternates. Occurrences shall be reported to the Head of the Department of Nuclear Engineering or designated alternates and to chartering or licensing authorities as required. Occurrences shall be reviewed by the Reactor Safety Board at its next meeting.
Reportable occurrences, including causes, probable consequences, corrective actions and measures to prevent recurrences, shall be reported to the NRC. Supplemental reports may be required to fully describe final resolution of the occurrences. In case or corrected or supplemental reports, an amended event report shall be completed and reference shall be made to the original report date.
- 36. ANSI-15.1, Section 6.8.3, "Records to be retained for the life of the facility," provides recommendations for record retention. TS 6.10.2 does not include copies of Reactor Safety Board Audits or records of review of violations of any safety limit, LSSS or LCO. Please propose changes to 6.10.2 to include these records or provide an explanation describing your reasons for not incorporating the changes.
Modify T.S. 6.10.2 to include the following items:
- j.
Reactor Safety Board Audits.
- k.
Records of review of violations of safety limit, LSSS or LCO
APPLICATION FOR LICENSE AMENDMENTS FOR THE TEXAS A&M UNIVERSITY AGN-201 TRAINING REACTOR Prepared by:
R. G. Cochran, P.E.
Professor and Head Nuclear Engineering Dept.
R.
J. Marzec Reactor Supervisor Nuclear Engineering Dept.
AUGUST 30, 1972
I INTRODUCTION This report is a request for license amendments for Texas A&M University's AGN-201 Training Reactor, Serial No.
- 106, U. S.
Atomic Energy Commission License No.
R-23 and Docket No.
50-59.
The license modifications requested are:
(1) Increase reactor power from 100 milliwatts to 5 watts.
(2)
Improvement and modernization of the AGN-201 reactor control system.
If the USAEC review and license modification can be accomplished by December 15, 1972, we would like to start the modification work January 1, 1973 and plan to complete the work on or before March 1, 1973.
The reactor power increase is being requested so that we can improve and expand many of our educational experiments performed with the reactor.
For example, the increase in neutron flux will permit more rapid measurements of neutron flux distributions and improved counting statistics.
The higher flux will also give improved data for control rod and safety rod calibrations.
The power increase will permit the inclusion of new experiments into our reactor laboratory courses that previously have not been feasible.
No increase in excess reactivity is being requested in connection with the requested 5 watt power level.
2 The planned modifications to the reactor control system will improve the operability of the reactor, and reduce the current maintenance problems.
Design Basis Accident Review A detailed safety analysis review has been conducted to ascertain the effects, if any, of power increase on the original design basis accident.
It has been concluded that the hazard to the reactor can only be associated with the temperature rise in the core and pressure buildup from fission product releases and hydrogen evolution from the polyethylene moderator.
An evaluation of these parameters using a 5 watt steady state initial condition and a 1% step change in reactivity (approxi-mately 150% of licensed excess) indicates that there would be no significant change in the hazard to the reactor, from that which presently exists with licensed operation at 100 milliwatts.
A summary of results obtained from our analysis of the excur-sion produced by a 1% step change in reactivity is presented in Table I.
The analysis was performed by use of a computer program which solved the point kinetics equation in finite difference.
The delay for rod drop was assumed to be 0.15 seconds.
A one second excursion time before shutdown was used in calculating the time dependent parameters of the reactor core.
In the case of
3 stuck rods, calculations for pressure buildup and total energy produced were based on the time to reach a center of core temperature of 1000C (core temperature rise 37.050 C) which is the expected melting point of the thermal fuse.
TABLE I.
Parameter Operating Rods Stuck Rods Max. power 2.66 x 103 watts 6.11 x 10 6 watts @.41 sec.
Max. neutron density 4.39 x 105 n/cc 9.97 x 10 8 n/cc @.41 sec.
Total energy 74.59 watt sec.
9.05 x 10 5 watt-sec.@2.61 sec.
Temperature rise
.003 0 C 37.050C @ 2.61 sec.
Pressure rise
.0023 psig 31.158 psig @ 2.61 sec.
The pressure buildup due to fission gases is on the order of 10-4 psig, which is negligible compared to the 31.158 psig total buildup after 2.61 seconds.
The bulk of the pressure buildup, would be due to hydrogen evolution from the polyethylene moderator.
The constants used in this calculation were for linear and branched polyethylene, while the polyethylene in the AGN-201 is reportedly cross-linked.
The actual pressure buildup from the hydrogen evolu-tion during the excursion would be expected to be somewhat less than that calculated because of the cross-linking bonds.
These results are consistent with calculations made several years ago by the Aerojet Corporation for the AGN-201 reactor.
4 Site The reactor site and principle activities carried on within the Exclusion Area have been defined previously.
Ref.:
Application for Amendment to Facility License No. R-23 for the AGN-201 Training Reactor at Texas A&M University dated October 12, 1970.
Exclusion Area The Exclusion Area, as defined in the 10 CFR 100.3 regulations, has been identified as the Reactor Room (61 Build-ing 518).
The criteria for the determination of the Exclusion Area (TID-14844) p. 5) is the limiting exposure to whole body of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
Radiation Shielding The radiation protection provided by the AGN-201 facility was analyzed to insure that the Exclusion Area has been properly defined and to predict the radiation levels which would exist during 5 watt operation and the Design Basis Accident.
The radiation levels for normal 5 watt operation were ob-tained by linearly extrapolating the readings recorded during operation at 100 mW.(I)
Figure 1 shows the isodose contour lines (1)These readings were obtained using a Victoreen 440 ionization chamber calibrated by our Health Physics Department.
5 within the reactor facility.
The values are those expected for operation at 5 watts.
Measurement of the gamma spectrum indicated an average energy of 1.2 MeV.
This information then allowed us to calculate the radiation levels beyond the concrete walls of the facility.
From figure 1 it is seen that a high radiation area exists at the surface of the reactor tank.
The width of this area, however, will be less than six inches.
We plan to consider any location within the reactor room, other than the shielded hallway, to be included in a
An audible and visual alarm will be installed to warn the reactor supervisor and personnel upon entrance to this area.
A radiation area will exist in the work area around the thermal column in the accelerator room.
Entrance from the accelerator room to the reactor room will be prevented by a floor grating around the thermal column tank; with this tank removed, entrance will be prevented by the installation of floor grating or closing and securing the accol-erator room door.
The radiation levels in both the control room and the M.E.
laboratory adjacent to the reactor room are extremely low, Shielding calculations for the design basis accident indicte that exposure due to prompt gamma-rays would be 11.29 R while that due to capture gammas will be 0.687 R.
Thus at the tank surface the total ex-posure during the accident due to gamma-rays will be 11.98 R.
The expo-sure in the control room, however, would be decreased to.022 mR, and the dose in the adjacent M.E.
laboratory will be.0366 mR.
Thus the area outside the reactor room itself would receive very small. doses.
Figure 1 Isodose Contours for operation of the AGN-201 at 5 Watts
- All readings are in mr/hr.
01
Figure 2 Isodose Contours for the accident of the AGN-201 while at 5 Watts All readings are in Roentgens.
8 Figure 2 indicates the isodose lines in the reactor room for the accident.
The exposure at the platform around the thermal column would be approximately 3.11 R.
We therefore conclude that no major changes in our radiation protection procedures are necessary for operation of the reactor at 5 watts.
The exclusion area will not need to be redefined as it is completely contained within the reactor vessel itself.
NUCLEAR INSTRUMENTATION The proposed Nuclear Instrumentation Systems for the AGN-201 will have the same number of channels as currently in use; however, an improved scram system and annunciator system is planned.
Our present reactor control system consists of the following:
CHANNEL #1 - BF3 counter with separate high voltage power supply, preamplifier, linear non-overloading amplifier, log cir-cuit, audio system and log n meter display.
CHANNEL #2
- Uncompensated ion chamber with separate high voltage power supply, and log n, period micro-micro amplifier (Keithley 420A).
CHANNEL #3 - Uncompensated ion chamber with separate high voltage power supply, and linear micro-micro ammeter (Keathley 410).
SKIRT MONITOR -
GM tube with an integrated power supply and ratemeter unit.
9 Channels 1, 2, 3, provide level data to individual "sensitrols" which provide high and low level signals for initiation of a scram by biasing a 6L6 tube to cut off.
The 6L6 vacuum tube provides current for the scram relay and the control rod magnets.
The Channel 2 period level is monitored by an electronic circuit in the scram chassis which initiates a scram by biasing the 6L6 tube to cut off on sensing a period ! 5 seconds.
The AGN-201 interlock system also provides a large negative bias to the 6L6 if any of several interlock swtiches are open.
The skirt monitor utilizes a relay, energized on a high level by a meter contact, to open the interlock system thus pro-viding a scram, and to turn on the evacuation horn.
THE PROPOSED CHANGES (1) CHANNEL #1 - Install meter contacts and relay to provide low level interlock signal only.
This interlock will prohibit the insertion of fuel rods if the source is out of the reactor.
(2)
CHANNEL #2 -
Change meter contact on Log N meter to provide lox,& level signal interlock.
The channel must be indicat-ing a shutdown flux level before rods may be inserted.
This will insure instrument operation.
10 Install a Keithley Electronic Trip Model 4103 to provide the high level scram signal.
Utilize the period meter contacts for fast period scram signal.
(3)
CHANNEL #3
- Change meter contact on linear micro-microammeter to provide low level signal interlock.
Install Keithley Electronic Trip Model 4103 to provide a high level scram signal.
(4)
Remove the present scram chassis which includes; sensitrols and resets for Channels 1, 2, and 3; interlock relay, period monitoring circuit, the 6L6 and its biasing circuit, the scram relay, and their associated indicator lights and switches.
(5)
Install the proposed new scram chassis which contains the main components of the proposed scram circuit and interlock circuits.
(6)
Install relay chassis on skirt door, for low tempera-ture and earthquake switch interlock and annunciator circuits.
(7)
Change miscellaneous switches and wiring.
The systems will then be as shown in figure 3, Nuclear Instru-mentation Systems.
Channel #1 Speaker Intlk Ann.
Intlk Ann.
Channel #2 Ann.
Scram Ann.
I Lw Intlk Linear eAnn.
Picoameter An Keitliley High Level Scram S4103 Hih&
Trip Ann.
H.V.
Ann.
~~SpeakerSra Scram Ratemeter High Level Evac.Horn Amp Meter IEvc.or Anni.
Figure 3 Nuclear Instrumentation Systems I-.
12 The Interlock System Ref: Fig.
4 The Interlock Relay is energized if the interlock circuit is complete.
The interlock circuit will be broken when any of the following take place:
- a.
A rod drive system plug is removed
- b.
Reactor tank temperature:-520 0 C
- c.
Earthquake switch opens
- d.
Shield water level low (t 7" from top of tank)
- e.
The relay chasis (control console) is removed
- f.
Channel #3 low level
- g.
Channel #2 low level
- h.
Channel #1 low level.
Suitable circuit connections are made to provide annunciator (light and bell) signals for items b, c, d, f,
g, h above, as well as alarms for high pressure reactor core tank, thermal column removed, and low voltage from Channels 2 and 3 high voltage power supplies.
The operation of Channel
- 2 or #3 is not seriously impaired by a reduction of ion chamber high voltage so a scram is not warranted and an alarm is justified.
A decrease in voltage on Channel #1 will so rapidly degenerate the signal that a low level signal will occur, which by de-energizing the interlock relay initiates a reactor scram.
SRo-d Drive lInterlock Rela
- Sys. Plugs hield Water An1 hn emp.
Sw.
Ann. 3 High Level arthquake Sw.
Ann.12 Channel 2 r
High Level Shield Water eelS.Ann.
19 4j Channel 2 vel Sw-.
An0l
- II Minimum Period elyCasis0 OJ nteloc
.w Skirt Monitor k
-4 I~ntrl~cC.)
High Level U
Channel 3 An1 U
ow Level Ann.10 Manual Scram S r 0 w Channel 2...........
wfW Level hannel 1I t Ann.
8 Rod Drop Sw.
Rod Vow Count Rate
[Coarse Rod
[S f
]Scram Relay S crA 7ý M g ne t.,
Reactor Control Safety Systems Figure 4
14 The Scram System: Ref: Fig. 4 Individual rod magnets will be energized whenever their associated scram relay is energized.
This will occur if the scram circuit is completed, i.e.
Interlock relay energized Channel #3 Keithley trip relay energized Channel #2 Keithley trip relay energized Channel #2 period relay de-energized Skirt monitor high level relay de-energized Manual scram switch not depressed Individual rod drop switch normal (closed)
A scram will occur, directly, by the opposite conditions of the above and indirectly, by causing the interlock system to open.
Of course, neutron flux data from Channels 1, 2, and 3 will remain available even with power off to the rest of the circuits.
The de-energizing of a scram relay will place a D.C. poten-tial by capacitor discharge of the appropriate polarity on the rod magnets to decrease the time it takes for the magnetic field to collapse.
The automatic carriage return circuit operation remains the same as at present and is initiated on any scram mode of operation.
The auto carriage return will not be initiated if an individual rod
15 is dropped, i.e.
the carriage will remain up and the position indicator will indicate its previous position.
Reinsertion of a dropped rod will require the appropriate carriages to be lowered to allow the magnet to pick up the rod.
The present sequence requirements for driving the safety rods in turn and then the coarse rod, has not been changed.
Annunciator An annunciator system will be installed which will provide additional visual presentation of the condition of the reactor systems.
The attached circuit diagrams indicate the circuit numbers of the annunciator system that will be energized for various con-ditions.
(See Figure 5 for annunciator panel backlighted window nomenclature.)
All circuits require 24 VDC signal to energize the annunciator.
A separate 24 VDC power supply is to be installed.
As a guide to the reactor safety system, the operation of each alarm circuit is described below.
Ann. #1 Manual Scram Depression of the manual scram button will open the NC contacts breaking the scram string thus de-energizing the scram relays for each rod.
The rods will drop.
The NO contacts will be momentarily made initiating the annunciation of the manual scram light.
The rod drop lights, Ann. #14, 15, 16, will also be lighted
MANUAL SCRAM HI PRESSURE CORE TANK LOW REACTOR TANK TEMPERATURE LOW LEVEL SHIELD WATER PERIOD SCRAM SCRAM HI LEVEL CHANNEL # 2 LOW LEVEL CHANNEL # 2 SAFETY ROD
- 2 DROPPED SAFETY ROD
- 2 ENERGIZED SCRAM HI LEVEL CHANNEL # 3 LOW LEVEL CHANNEL # 3 COARSE ROD DROPPED COARSE ROD ENERGIZED SCRAM HI LEVEL SKIRT MONITOR LOW VOLTAGE SCRAM BUS TROUBLE CHANNEL # 2 H.V. SUPPLY TROUBLE CHANNEL # 3 H.V. SUPPLY INTERLOCK OPEN EARTHQUAKE SWITCH OPEN ACCELERATOR DOOR OPEN THERMAL COLUMN REMOVED ROD HOLD SAFETY ROD
- 1 DROPPED SAFETY ROD A#I ENERGIZED Annunciator Window Nomenclature Figure 5
17 by action of the scram relays.
Acknowledgement of the alarm will reset the manual scram light.
Ann. #2 Period Scram The period scram set point is controlled by the movable red pointer contact on the period meter of Channel #2.
A fast period that causes the meter (black) pointer to reach the red pointer make a circuit to energize an auxiliary relay in the Channel #2 chassis (See Figure 7 period circuit modification).
The meter relay will remain energized until the signal is below the set point.
When energized, the meter relay contacts will open the scram string causing the rods to drop and initiate the annunciator circuit causing the period scram light to turn on.
Rod drop lights Ann. #14, 15, and 16 will also light by action of the scram relays.
Ann. #3 Scram High Level Channel #2 The output of the Log N amplifier will be monitored by a Keithley Electronic Trip Model 4103 set nominally at 150% of licensed power, i.e.
7.5 watts.
The Keithley Model 4103 is specifically designed for use with the 400 series micro-micro ammeter.
It is completely self-contained and operates on 117 vac.
It represents the current state of technology in electronic trip units and has been previously reviewed by the USAEC for these appli-cations.
18 The Keithley Trip unit will be mounted on the amplifier chassis and will provide contacts for the scram string and the annunciator.
The trip relay is normally energized and will fail safe. (See Figures 8 and 9).
In the event of a high level (>150% of full power) the trip relay will de-energize opening the scram string and dropping the rods.
The annunciator circuit will be closed initiating the scram high level alarm and light.
The rod drop lights Ann.
14, 15, and 16 will also light.
Ann. #4 Scram High Level Channel #3 The output of the linear amplifier will be monitored by a Keithley Electronic Trip Model 4103 set at 95% full scale deflection.
All other components are the same as for Ann. #3.
Ann. #5 Scram High Level Skirt Monitor The skirt monitor is a Geiger Mueller gamma-ray detector which provides an additional safety channel scram.
A meter contact set at 200% of licensed power (10.0 watts) will energize an auxiliary relay which will:
- a.
Open the scram string, dropping the rods (giving Rod Drop Annunciation #14, 15, and 16)
- b.
Light Ann. #5
- c.
Energize evacuation horns on ground floor and first floor of the Nuclear Engineering area.
19 Ann. #6 Interlock Open The interlock relay when de-energized (ref. page No.
12 )
will cause the scram string to open, rod drop annunciation and light the Interlock Open panel.
Ann. #7 High Pressure Core Tank The pressure switch will be connected to the core tank vent valve and set at 5 psi to initiate an alarm.
The system will not cause an automatic scram, but will require opera-tor action as described in the emergency procedures.
Ann. #8 Low Count Rate Channel #1 A meter contact will be set at a low count rate to insure that the system is operating and that the neutron source is in its'proper position in the reflector.
A count rate less than that of the meter contact will cause the interlock circuit to open, and the following lights will turn on:
Low level Channel #1 Interlock open Safety Rod #1 dropped Safety Rod #2 dropped Coarse Rod dropped
20 Ann.
- 9 Low Level Channel #2 A meter contact set at approximately I x 10-12 amps on the micro-micro-ammeter (based on sensitivity range of the UIC) will energize an auxiliary relay on the low level or loss of signal to open the interlock circuit, de-energizing the interlock relay and causing a scram.
The following lights will turn on:
Low level Channel #2 Interlock open Safety Rod #1 dropped Safety Rod #2 dropped Coarse Rod dropped Ann. #10 Low Level Channel #3 A meter contact set at 5% of full scale deflection will energize an auxiliary relay on a low level or loss of signal to open the interlock circuit, de-energizing the interlock relay and causing a scram.
The following lights will go on:
Low level Channel #3 Interlock open Safety Rod #1 dropped Safety Rod #2 dropped Coarse Rod dropped Ann.
- 11 Low Voltage Scram Bus A relay will monitor the 28 volt scram and relay bus voltage de-energize if the scram bus becomes overloaded or otherwise de-creases below an appropriate setting.
(This setting will be determined when the new circuit nominal load is established).
21 Ann.
- 12 Earthquake Switch Open The earthquake switch is a gold plated ball which rests in a small detent.
Horizontal forces will cause the ball to move momentarily out of the detent, opening the circuit to an auxiliary relay.
The relay provides continuity for the interlock system and when de-energized, will scram the reactor and annuci-ate these conditions:
Earthquake switch open Interlock open Safety Rod #1 dropped Safety Rod #2 dropped Coarse Rod dropped Ann.
- 13 Low Reactor Tank Temperature The low temperature switch is set to open at
- 20 0 C.
It controls an auxiliary relay which when de-energized, will open the interlock system and annunciate these conditions:
Low reactor tank temperature Interlock open Safety Rod #1 dropped Safety Rod #2 dropped Coarse Rod dropped
22 Ann. #14 Safety Rod #1 Dropped Ann. #15 Safety Rod #2 Dropped Ann. #16 Coarse Rod Dropped The scram relays for each rod when de-energized, will initiate an alarm and turn on the appropriate panel light.
Ann.
- 17 Trouble Channel #2 H.V. Supply Ann. #23 Trouble Channel #3 H.V. Supply A relay will monitor each power supply and de-energize when the voltage decreases substantially.
The operation of the UIC is essentially not degraded unless a large decrease in voltage occurs.
The relays initiate the applicable alarm and annunciator light when de-energized.
Ann. #18 Accelerator Door Open A limit switch will initiate an alarm and light when the accelerator door is opened.
If the alarm is acknowledged, the light will remain on as long as the door is open.
Ann. #19 Low Level Shield Water An auxiliary relay will be controlled by the float switch in the shield water tank.
If the level of water in the shield tank becomes t-:-7 inches from the top of the tank, the relay will
23 de-energize, opening the interlock string and de-energizing the interlock relay.
The reactor will scram and the following lights will go on.
Low Level Shield Tank Interlock Open Safety Rod #1 dropped Safety Rod #2 dropped Coarse Rod dropped Ann. #20 Safety Rod #1 Energized Ann.
- 21 Safety Rod #2 Energized Ann. #22 Coarse Rod Energized When there is continuity in both the interlock and scram systems, the scram relays will be energized and the annunciator will indicate that the rod magnets are energized.
Ann. #24 Thermal Column Removed Rod Hold A microswitch opens when the thermal column is removed, which interrupts the rod drive system, preventing rod motion.
An annunciation of this condition is also provided.
24 MODIFICATIONS TO THE PRESENT ADMINISTRATIVE CONTROLS AND EMERGENCY PROCEDURES Operation at 5 Watts The reactor supervisor will directly control access to the reactor room.
He will insure that only those persons necessary to accomplish the experiment are authorized to enter the reactor room while the reactor is approaching or operating at 5 watts.
Up-on opening the reactor room door an alarm and flashing red light will be activated.
Likewise, access to the accelerator room, which is above the reactor, will be controlled by a member of the Nuclear Engineering Department staff when the reactor is to be operated at 5 watts, and the accelerator door is open.
The reactor supervisor will be in direct communication with the accelerator room and will receive an annunciator alarm and light if the accelerator room door is opened.
High Pressure Core Tank Alarm -
Procedure at. or Approaching 5 Watt Operation:
Manually scram, observe and record all flux, temperature, and radiation levels.
Verify the alarmed condition by observing the pressure indicator.
If pressure reading is abnormal, notify the head of the Nuclear Engineering Department or his designated alter-nate.
25 TECHNICAL SPECIFICATIONS The present technical specifications contained in Appendix A to the license require that all safety channels "be operating whenever any control or safety rod is not in its fully withdrawn position, except that either safety channel 1 or 3 may be by-passed" (refer to paragraph 3.3, page 2 of the Technical Speci-fications which you have).
Furthermore, Table I, page 6 of the Tech Specs, requires that channels 1, 2, and 3 have both high power and low power scram functions.
The previously described operation of the proposed scram and interlock systems described in this report, when approved, will necessitate a change in these areas of the specifications to reflect the deletion of the high power scram on Channel #1 and the deletion of the reference to sensitrols in the column headed, "Limiting Safety System Setting" in Table I.
We suggest that the following wording be used:
(page 2 para. 3.3) each of the safety channels in Table I shall be operating whenever any control or safety rod is not in its fully withdrawn position.
Each of the operating safety channels shall sound an alarm and cause automatic reactor shutdown if the limiting safety system setting is reached.
26 (page 6, Table 1)
TABLE I NUCLEAR INSTRUMENTATION Function Safety Channel Limiting Safety System Setting Low count rate Low power High Power Short Reactor Period Nuclear Safety No.
I Nuclear Safety Nos.
Nuclear Safety Nos.
2&3 2&3 Loss of 200% of power 10 cps signal licensed Nuclear Safety No.
2 5 second minimum period