ML20209F176

Forwards Partial Response to 981020 RAI Re GE Nuclear Test Reactor Sar.Reply to Each Question Follows Italicized Restatement of Question/Comment.Remaining Unanswered Questions Will Be Addressed in Later Submittal

ML20209F176
Person / Time
Site:	Vallecitos Nuclear Center
Issue date:	06/18/1999
From:	Murray B GENERAL ELECTRIC CO.
To:	Mendonca M NRC (Affiliation Not Assigned)
References
NUDOCS 9907150250
Download: ML20209F176 (20)

Text

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GENucIcarEnergy Geres!Deant Campsy Vallocnos Nuclear Centn 6705 Vallecaos Ruud. Suno!, CA !M586 June 18,1999 Marvin M. Mendonca, Senior Project Manager Non-Power Reactors and Decommissioning Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washmgton, D.C. 20555-0001'

References:

1. Docket Number 50-73, License R-33.

2. " Request for Additional Information (TAC No. MA0099)", Marvin M. Mendonca to G. L. Stimmell, dated October 20,1998,

3. NEDO-32740," General Electric Nuclear Test Reactor Safety Analysis Report",

August,1997.

Dear Mr. Mendonca:

I A partial response to the request for additional information (Reference 2) regarding the General Electric Nuclear Test Reactor Safety Analysis Report (NTR SAR, Reference 3) is submitted in the attachment to this letter. The reply to each question follows an italicized restatement of the question / comment. The remaining unanswered questions will be addressed in a later submittal.

If there are additional questions related to this response, please contact me at (925) 862-4455.

Sincerely,

~P ep ,

i B. M. Murray, Senior Licensing Engmeer Regulatory Compliance' +

Attachment cc: Mr. Steve lisu ,

Radiologic liealth Branch ,

t State Department of ficalth Service P.O. Box 942732 Sacramento, CA 94234-7320 9907150250 990618 l PDR ADOCK 05000073 t

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l General Ouestion No.1: In applicable chapters ofthe SAR, describe the interlocks and permissive circuits related to the NTR that are not already discussed in the SAR. Provide or reference an analysis that demonstrates the adequacy to achieve designfunctions of all NTR-related interlocks orpermissives without a TechnicalSpecification (TS) requirement.

l Alternatively, provide TS requirementsfor the applicable interlocks orpermissives.

There are a number ofinterlocks and permissive or bypass switches at the NTR. These are listed and discussed below;

a. Radiation Monitor / Stack Gas llorn Bvnass This bypass switch allows bypass of the Control Room horn after the alann is received. The bypass switch is located on the reactor console under the stack effluent recorder. The switch activates an adjacent red light when the switch is in the bypass position to prevent inadvertently leaving the switch in the bypass position. Also located on the same panel are two yellow lights which illuminate when a high radiation monitor alarm or high stack gas activity condition exists.

The bypass switch is normally left in the unbypassed position. The Radiation Monitors are described in Section 7.6. The Stack Monitoring System is described in Sections 6.3 and 7.4.

b. North Shutter Locic and Override The north shutter is interlocked with the reactor cell inner door. When the reactor cell inner door is open, the north shutter will close and not be able to be opened. This is to protect personnel entering the reactor cell from being exposed to radiation beam from the north room radiography area. The only radiation beam would be from radioactive items left in the radiography position.

The override allows the shutter to be opened for maintenance after verifying there are no radioactive objects in the north radiography position. The logic override switch is a keylock switch with red and green lights indicating switch position. The switch is normally left in the non-ovenide (green light) position. The switch and associated indicator lights are located on the shutter control panel on the reactor console. The north room radiography area is described in Sections 1.3,4.3,10.2, and Figures 1-1 and 6-2.

c Rod Selector Switch The rod selector switch is a keylock switch on the reactor console with a red light which illuminates when the selector switch is armed. The purpose of the switch is to allow Control Rods or Safety Rods to be withdrawn out of sequence for maintenance and testing purposes.

During maintenance or testing, the keylock switch is armed and the red light illuminates. A separate switch is then rotated to select a Control Rod or Safety Rod, and then the selected rod is withdrawn. All other rods must be fully inserted in order to withdraw the selected rod. The rod selector switches and indicator light are located in the reactor console. The rods and the test panel are discussed in Sections 4.2.2 and 7.1.

d. South Cell Interlock ud Iktn Bynass l The South Cell door interlock system is mentioned in Sections 4.3,4.4.1 and 7.6. A photocell across the door entrance gives an audible alarm upon entry. A visual light also illuminates when the South Cell dooris opened.

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• I An interlock has been established between the door, the shutter and the radiation area monitor, f

When the radiation level exceeds the trip point, air to the door operator is cut off to prevent the i door from being opened. If the door were to open, an audible and visual alarm would be activated. Similarly, if the radiation level at the radiation area monitor were to increase above the set point of the monitor while the door were open, an audible and visual alarm would be )

activated, and the shutter would automatically close. The horn bypass switch bypasses the audible portion of the alarm. All interlocks and automatic features to protect individuals from exposure to the beam are still functional. The bypass allows the South Cell door to be opened past the door switch open position for instrument cables to enter the cell during special instrumented experiments. Operating in this mode is only done in accordance with a written procedure, reviewed by the independent review function and approved by the Facility Manager.

The bypass switch is located on the reactor console.

e. South Cell Photocell Bvnass Switch As mentioned above, a photocell sensor across the South Cell doonvay activates an audible alarm when personnel enter the cell. A bypass switch is mounted on the reactor console which silences the alarm. This bypass is currently not in use and the alarm remains in the operable position.

f. Low Flow Scram Bvnass and Test Switch l The NTR low prinnry coolant flow scram is only required when the reactor is operating above 100 watts. Below that power level, this scram is bypassed as described in Section 7.3.2. The bypass switch is described in Sections 5.2 and 7.3. The bypass switch is located in the Log power instrument which is located in the reactor console. When the Log power instrument trips are reset, the bypass switch is reset if the instrument indicated power is below 100 watts. As a consequence, when the reactor is shut down and the Log power instrument is reset, the low flow scram bypass is in the bypass position. The low flow scram cannot be tested. The test switch allows the scram function to be tested. This push button, momentary switch bypasses the bypass and allows the scram to be activated when the primary coolant pump is turned off.

g. Manual Override on Shutter Timers There are two shutter timers for each shutter. One timer is automatically reset to a preestablished open time. The other timer must be set each time the shutter is opened. An override switch bypasses the timers and allmvs the shutters to be opened manually for an indefinite time period.

This is used during radiation surveys and for experiments requiring an indefinite exposure to the beam. The manual switches are located on the reactor console and do not override any of the shutter interlocks.

h. Macnet Test Jack A magnet test jack is located on the reactor console. Thisjack allows test leads from the magnet L current power supply to be used for Safety Rod magnet testing during shutdown maintenance. A keylock switch switches the magnet current from Safety Rod 3 magnet to the testjack. With the switch in the test position Safety Rod 3 does not have current sui y and, therefore, cannot be made up during shutdown and would cause a reactor scram if th< . witch were activated during reactor operation, i

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i. Delta Temperature Automatic Switd) ,

A thermopile measuring the delta temperature on the reactor core inlet and outlet coolant pipes shares a recorder with the Source Range Monitor (SRM) count rate. An automatic feature switches input to the recorder from the SRM count rate to the delta temperature. This occurs at the Low Power Scram Flow Bypass Setting if the switch did not occur manually when the reactor critical position is verified on startup.

j. Rods in Prior To Scram Reset All Control and Safety Rods must be fully inserted before the safety system can be reset to remove any rods. This interlock is described in Section 7.

k. Rod Scauential Removal An interlock requires each Safety Rod to be removed in sequence and requires all four Safety Rods to be fully removed before any Control Rods may be removed. This interlock is described in Section 7.

1. Picoammeter Trins and Bypass.

The picoammeter high-power trip two-out-of-three coincidence logic and the momentary bypass switches are discussed in Section 7.3.3.

m. Rod Block The picoammeters contain a downscale alarm to indicate meter readings which may not be adequate to properly monitor neutron level increases. When two of three picoammeter meter readings are below the set point, a block occurs which prevents Safety Rod or Control Rod removal. This is described in Section 7.4.

n. Control Rod and Reactor Source Automatic Insertion Upon receiving a process scram, the reactor Control Rods and the reactor source automatically l run to the full-in position. This is described in Sections 4.2.2,4.2.4 and 7.7.

l l General Ouestion No. 2: Abnormal, offnormal, unusual event or conditions, anticipated

! operational occurrences and afew other terms are used in the SAR without apparent distinctions between these terms. Provide a consistent use ofterms in the SAR.

l Anticipated Operational Occurrences are specific events which an 2 scribed in Section 13.3.

This term occurs on pages 1-2,13-1 (four times),13-2 and 13-56. It should be noted that the l l '

term " anticipated abnormal occurrence" appearing on pages 4-22,13-56 and 13-58 should be

" anticipated operational occurrence". All other words, such as abnormal, off-normal or unusual, ,

I or terms using these words are used in their dictionary defmition or as they are commonly used.

These words are not uniquely defined in the SAR.

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I Pane 1-2: Provide the reasons thatpotential human error was not included as a postulated initiating event or demonstrate that the considered events boundpotential human error type events. i l

u Potential hitman error is included as a source of anticipated operational occurrences and accidents. For instance, a fuel handling error or an inadvertent starting of the primary pump '

would be caused by human error. Other anticipated operational occurrences, such as rod  !

withdrawal, could be caused by human error or equipment malfunction. The anticipated i operational occunences focus on postulated credible events and less probable accidents and I subsequent consequences without specifically listing the cause of the event. All human error l type events are included in those listed. I Pane 1-7: Current NRClicenses include the VBWR, GETR, and the EVESR, although section 1.8 indicates these licensed activities had end dates. Provide clarification.

Tne list on this page was provided as support for " experience at the site with NRC licensed activities...". GE does currently hold Possess-Only Licenses for VBWR, GETR and EVESR.

The dates on page 1-7 reflect the experience accumulating on operating reactors.

Paec 1-10: Provide discussion of thepurpose ofthis letter and Appendix A to it.

The letter and Appendix A to the letter are the agreement discussed in Section 1.7 and support that information.

Paec 2-5: Provide additional descriptionfor title " Safety. "

This section titled " Safety" appears in Section 2.1.1, " Site Location and Description." The brief discussion on safety relates to safety pertaining to the site location on the west coast of Northern California and in the Vallecitos Valley. The section could be titled " Site Location Safety" or l " Weather Hazards."

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Pane 2-9: The secondparagraph in Section 2.1.2 " Population Distribution" indicates that

"(t)here are approximately ten houses immediately west of the site." Provide a quantitative l indication of the distance ofthe nearest to the NTR. Verify that these are within the 1-Am

. radius or not. If they are within the 1-km radius, provide clanfication on thefirst sentence of this paragraph that implies there are onlyfour houses and one BMX race track.

On the south side of Vallecitos Road, there is the BMX bicycle racetrack and one occupied house within the 1-km radius. An additional unoccupied property appears to contain an actively used i barn and is used for storage of farm equipment. Adjacent to the site on the west are eight houses )

within 1 k.m and live houses exceeding the 1-km radius of NTR. Therefore, there are nine houses i

1 and one BMX bicycle racetrack within the 1-km radius of NTR. The closest houses to the NTR were measured at 2,400 feet. These houses are 1) the house south of Vallecitos Road adjacent to )

the BMX racetrack and 2) the second house from Vallecitos Road on Little Valley Road which adjoins the site to the west.

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Paec 2-11: As described earlier in the Safety Analysis Report, the site is surrounded by barren mountains and rolling hills. Provide an analytical description ofthe effects ofthis terrain on the effluent releasesfrom the stack in the meteorology section or other appropriate section.

1 (To be provided later)

Page 2-11:

Provide conclusionsfor the meteorology ard hydrologv sections 2.3 and 2.4, respectively.

l The meteorology, hydrology, and seismology sections refer to documents that are not readily k '

l apparent or available. Provide these documents or appropriateportions ofthese documents, orprovide readily available reference to these documents.

l Reference to these documents was inadvertently deleted from this revision of the SAR. The )

l references and conclusions are listed below and were used in the license renewal application j (revised SAR and Technical Specifications) for NTR, License R-33, Docket 50-73, submitted i April 29,1981:

1. Environmental Information Report fbr the General Electric Test Reacto_r, General Electric Company, Vallecitos Nuclear Center; Pleasanton, California; July,1976 (NEDO-12623).

a. Water-Ouality Monitorine Network for Vallecitos Valley. Alameda County. California, U.S. Geological Survey, Water Resources Investigations 80-59.

2. Geolocic Investication. Phase II. General Electric Test Reactor Site. Vallecitos. Califomia; Earth Sciences Associates, prepared for General Electric Company, Pleasanton, California; February,1979.

3. Seismic Criteria and Basis for Structural Analysis of Reactor Buildine. Engineering Decision Analysis Company, Inc., prepared for General Electric Company (GETR), Pleasanton, California; December,1977.

4. Probability Analysis of Surface Runture Offset Beneath Reactor Buildine. General Electric Test Reactor, Engineering Decision Analysis Company, Inc., prepared for General Electric Company, San Jose, California; April,1979 (EDAC 117-217.13).

5. 'L. Kovach, A Seismolonical Assessment of the Probable Exnectation of Strone Ground Motion at the General Electric Test Reactor Site; April,1980.

6. Review of Seismic Design Criteria for the GETR Site, Engineering Decision Analysis Company, Inc., prepared for the General Electric Company, San Jose, California; April, 1980.

Paec 3-1: Provide an analysis that dernonstrates the adequacy to maintain personnel 1 radiation exposure less than 10 CFR Part 20 limits of the beam shutters interlocks without a l TechnicalSpecification (TS) requirement. Alternatively, provide TS requirements to the beam l

shutters interlocks. \

Facility shielding has been provided which maintains personnel radiation exposure less than 10CFR20 limits. .This shielding continues to be improved. Additional shielding was added in J 1997. There are some posted areas where a person could receive exposures in excess of 10CFR20 limits. These areas, however, are posted as a Radiation Area or High Radiation Area as appropriate, and access to these areas is restricted as required by the regulations. The South Cell is an area that has additional access restrictions because of the frequency of entrance and the potential for inadycrtent entrance. These additional restrictions include the alarms and interlocks described in the SAR. These alarms and interlocks are designed to prevent unnecessary exposure and minimize the probability of an exposure which would exceed 10CFR20 limits. The remote gamma radiation monitor located in the South Cell indicates a dose rate of 600 mrem /hr on the wall adjacent to the exposed beam. Short term exposure to the beam in the South Cell with the shutter open (a typical neutrograph exposure is four minutes) will not likely result in an exposure exceeding 10CFR20 limits. The exposure would be unnecessary, and, therefore, extra precautions have been taken to prevent entrance with the shutter open. The controls maintain exposures ALARA, therefore a Tech Spec is not required for this interlock.

bge 3-2: Provide clarification on the use ofthe term reactor confinement building in the firstparagraph.

Reactor Confinement Building, as used here, conforms to the ANSI /ANS-15.1 definition as a

" closure on the overall facility which controls the movement of air into it and out through a controlled path." Building refers to the cell which contains the reactor core and associated water and control systems and is contiguous to the rest of the facility such as South Cell, Control Room and North Room.

I l Paec 3-2: Provide an analysis ofthe potential compaction ofthefuel withoutprimary water loss.

As indicated in Sections 3.4 and 13.1, compaction of the fuel would result in a water loss. Fuel compaction is postulated, although extremely unlikely, from external forces. Since the reactor I

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1 f core is contained within a block of grechite, there is no possibility of compacting the core without resulting in breakage of the core can and primary coolant piping which would result in  !

water loss. Therefore, the occurrence of a compacted core with no water loss is not credible and i has not been analyzed.

Paec 3-2: Provide an analysis to verify that the cadmium poison sheets "willnot more relative to the core during a seismic event."

The Manual Poison Sheets (MPS) contain a latch pin that rotates by an insertion tool and is spring return to the vertical position. When installing the MPS, the latch pin is rotated with the tool and inserted until the pin touches a guide block. The tool is then relaxed and the pin rotates under a latch plate. The MPS is then pulled to verify that it is latched and cannot be pulled out (or ejected by seismic forces). The guide plate and latch block are attached to the graphite guide block which is an integral part of the graphite block assembly supporting the core. The measured movement relative to the core is approximately 1/16 inch, but varies depending on manufacturing tolerances. The latch pin will rotate radially to latch and unlatch the MPS, but there is no axial movement of the pin. The pin, then, does not contribute to movement of the MPS relative to the Core.

Pane 3-3: Provide the applicable reference (s)for the seismic structural analysis.

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The seismic structural analysis may be found as an Attachment A to the letter, R.W. Darmitzel to E.G. Case, dated November 29,1977 (License R-33, Docket 50-73). Additional information is included in the letter, G.E. Cunningham to D.C. Diianni, dated October 18,1979 (License R-33, Docket 50-73).

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Paec 3-3: Provide a description ofthe administrative and system controls, interlocks and TS requirementsfor the reactor cell and stack ventilation system that are required to protect against potentialfueled experimentfailure.

To protect against a potential fueled experiment failure, the following controls are provided: l

• Technical Specification requirement for the reactor cell to be negative pressure in order to operate the reactor above 100 watts e Technical Specification requirement to limit the amount of radioactive material discharged from the reactor stack e SOP requirement to DOP test the ventilation system IIEPA filters annually

• SOP requirement on the reactor cell flow rate i

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! l 1%e 4-1: Since the current core " container was put into service in 1976 after the previous container, which had been in servicefor approximately 18 years, sprung a leak in a weld i area," provide an analysis ofthe current condition ofthe container. Includepotential weld and material neutron embrittlement considerations. Also, according to the discussion on page 4-10 the configuration allows inspection ofthefuelcontainer. Provide the results ofthese mspections. In addition, provide reasoningfor a TS requirement onfrequencyforfuture inspections, i

In this regard, please docuss,from an agingperspective, the current condition ofother safety related reactor components such as control and safety rods and their drive mechanisms, wiring, coolant piping and other safety related wiring and relays, including relay contacts. If I inspections of these components have been performed, please provide the results.

J (To be supplied later.) )

i Paec 4-6: Provide indication whether the safety and control rods arepowder or solid baron carbide material. Ifthe rods arepowder, discussfeatures or controls that ensure detection of potentialrod swelling.

The Safety and Control Rods are solid boron carbide material. Each Safety Rod is made up of ten %-inch by 2-inch boron carbide cylinders, and each Control Rod is made up of eight %-inch by 2-inch cylinders.

Pane 4-10: The cable held retractable irradiation system was not observed in Figure 1-1 or in Chapter 10. Provide an appropriate drawing of this system.

In the discussion of the CIIRIS facility, first paragraph: "The tube runs through the NTR reactor cell wall, across the reactor cell, and connects to the experiment position". This experiment position can be seen on Figure 4-1, denoted as "S" experiment tube.

Paec 4-12: Provide thefiguresfor Chapter 10 that are referenced under thefirstparagraph of the " Reactor Shield"section. Similarly,for the referenced Chapter 10 figure in the " North Room Modular Stone Monument"section on page 4-14. Also, pages 10-1 through 10-8 rtference drawingsfor the experimentalfacilities, but they were not included in Chapter 10.

Page A-4 in Appendix A also refers w Chapter 10 figures.

The figures in Chapter 10 were inadvertently left out of the document. These figures are attached as Figures 10-1 through 10-5.

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4 Pane 4-15: The maxitnum pressure specified in the core may be inconsistent with the description of an unpressurizedprimary coolant system on page 5-1. Provide clarification.

- The reactor core is lower than the top of the water level in the connected fuel loading tank. The height of the water level in the tank causes a static water pressure of 20 psia at the core.

Pane 4-15: Potentialexcess reactivity"is definedin the TS. Provide a ddinition whenfirst used in the SAR or reference the TS definition.

On page 4-15, fourth paragraph, first sentence should reference Technical Specification 1.2.15.

Pane 4-23: Regarding shutdown margin, TS 4.1.3.3 requires " calculation or measurement whenever a decrease in the reactivity worth of a safety rod is suspected. " Provide description ,

ofconditions when this calculation or measurement is required.

A decrease in the reactivity worth of a Safety Rod would be suspected if, during reactor startup, there were an unexpected increase in the suberitical multiplication, or the reactor were to go critical before fully removing the four Safety Rods.

Additionally, a major change in the reactor core, the reactor control system, Manual Poison Sheet configuration or fuel loading would also indicate a potential decrease in the reactivity worth of a Safety Rod.

Paec 4-25: Provide reference 7for the Jens-Lottes correlation.

Reference 7 is W.H. McAdams, Heat Transmission, third edition, McGraw-Hill, New York, 1954, pp. 362 and 393.

l Pane 4-25: Providefurther detailon the derivation ofthepeakingfactors. Provide a description or referencefor what combination ofexperimental and analytical methods to i derive these peakingfactors. Similarly, provide an explanation of"with neutronfluxpeaked l t

on one side of the core. " Describe the measurements and/or analyses used to establish circumferential and axialpower proflies. Compare the results of these derivations to the assumptions used in this safety analysis report.

I (To be provided later.)

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Pane 5-4: Regarding the allowablepr: mary system leakage rate of10 gallons / day, provide or reference, a calculation ofthe maximum radiological aposure to radiological workers and to thepublic considering the maximum allowable radioactivity in the primary coolant system.

Also, a similar calculation should be madefor the potentialloss ofcoolant accident analysis in Chapter 13. The radioactia source term should be consistent with ithe statement on page 5-9 that "(a)n inadvertent release ofexcess radioactivity in the primary coolant, ofhigh enough level, would cause the reactor cell remote area monitor to alarm. " That is, the level of radioactivity in the primary should be thatjust below what would initiate this alarm, or at a level corresponding to other indications that would terminatefacility operation and initiate corrective actions.

(To be provided later.)

Pane 5-5: Provide clanfication on the discussion ofthe primary coolant high core outlet temperature <200 F. If this is the alarm that is discussed in the nextparagraph, provide indication ofsuch.

The primary coolant high core outlet temperature 200 F is the maximum set point for the alarm.

Pane 5-5: Provide clanfication or reference to the use of the 222 *F value in the safety analyses to ensure nofueldamage.

The critical heat flux is a strong ftmetion of power but is not significantly affected by core inlet temperature. The safety limit for reactor power assures that the actual heat flux never approaches the departure from nucleate boiling heat flux (reference page 13-49). This protects the fuel from damage. Additionally, the loss-of-flow accident without reactor scram (13.4.5) and the loss-of-primary-coolant accident without reactor scram (13.4.6) conclude there is no fuel damage from these accidents. The reactor temperature scram, then, was conservatively included and the set point arbitrarily conservatively set at 222 F.

Paec 5-9: Provide the analysis that shows that the 10' mR/hr alarm provides acceptable personnel and reactorprotection. Include any other systems or alarms that ensure detection and correction ofinadvertent release ofradioactivity in the primary.

The area monitor in the reactor cell provides an indication of an inadvertent release of excess radioactivity in the primary coolant system. This release would come from the reactor fuel since no other source of radioactivity release is present in the primary coolant system or the core. A l significant release is not likely since the reactor fuel is not damaged in the event of anticipated operational occurrences or postulated accidents, and aluminum uranium alloy is not prone to I release large amounts of radioactivity unless there is severe damage to the fuel. The monitor set l

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point was established to be higher than the operating level to prevent spurious alanns, yet not high enough to preclude alarm notification of an increase in radioactivity level. During reactor shutdown periods when work is being performed in the reactor cell, the monitor set point is reset to 100 mrem /hr as noted in Table 7-2. A control room radiation area monitor indicates the radiation level in the control room and protects personnel during reactor operation. Other systems that detect inadvertent releases of radioactivity to the primary coolant system are the stack effluent monitor (the primary coolant system is open to the reactor cell atmosphere through the fuel loading tank), conductivity monitor, primary coolant sampling and other area radiation monitors outside the reactor cell.

Pane 5-9&10: Provide or reference the analyses that demonstrate that the specification on primary coolant water conductivity ensures aluminum corrosion is within acceptable levels.

(To be provided later.)

Paec 6-1:

Provide the configuration ofthe controlrods during refueling. Also, discuss theposition of any safety rods that are not inserted to satisfy minimum shutdown margin requirements.

Since the position of the reactor cell door during refueling appears to be optional, provide an analysis to demonstrate that it need not be maintained closed. l It has been more than 20 years since the reactor fuel was removed. It is not expected to move the fuel for an additional decade. Consequently, the NTR does not attempt to maintain current refueling procedures. Individual defueling and refueling procedures will be developed (old procedures modified) and approved in accordance with the Technical Specifications when they are needed.

Paec 6-5: Provide information on the stack configuration (e.g., height, diameter, draft relacity, close by buildings, delta y, effect ofhilly terrain).

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' he reactor cell exhaust stack is approximately 14 metets (45 feet) above grade level of Building j

105 (Tech Spec 5.1.5) and approximately 2.7 meters (9 feet) above the highest portion of the l Building 105 roof. The maximum stack flow rate with the reactor cell door closed is nominally 1,800 cubic feet per minute. This flow rate is equivalent to an average air flow velocity in the stack of 1,420 feet per minute. The stack is 13 % inches square inside dimensions by 18 feet high located on top of the reactor cell.

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i The reactor cell AP, with the door closed, is approximately 2.0 inches of water. The cell negative ,

pressure is verified prior to the first reactor startup each day (Tech Spee 4.4), and reactor power shall not be increased above 100 watts unless the reactor cell is maintained at a negative pressure of not less than 0.5 inch of water whh respect to the control room (Tech Spec 3.4.3.1).

Figures 2.3 and 2.4 show there are no other buildings near Building 105. The location of the NTR, near the center of the occupied area of the site, results in a surrounding terrain that is gently sloping or rolling and has no deleterious effects on effluent releases. The annual average dilution-dispersion determinations and accident evaluati .is conservatively use a ground release i assumption. j Pane 6-6: Provide more specific reference to discussion ofstack sampling / monitoring system in Appendix A.

Sections 6.3 and 6.4 provide sufficient details for the stack sampling / monitoring system. The reference to Appendix A is in error.

Pune 7-1: Thefirstparagraph and Figure 7-1 indicate a high tog Npower trip, but Table 7-1,  !

Section 7.3.3 and the TechnicalSpecifications do not appear to discuss this trip. Another trip j that appears to be only in the Figure is thatfor seismic disturbance. Provide clarification or  :

reference to another section ofthe SARfor which reactor safety system scrams, interlocks or other instrument and control systems and components are requiredfor the accident analyses  !

and by TechnicalSpecsfication. Also, provide clanfication or reference to another section of l the SAR on which scram, interlocks or other instrument and controlsystems and components provide additionalprotectionfunctions that are not specifically used to mitigate accident analyses or are required. A table may be appropriatefor this clanfication. Further, provide verification that the TechnicalSpecifications Tables 3-1 and 3-2 are consistent with this safety analysis.

The Log N high-power trip and the seismic scram are not required to ensure the safe operation and shutdown of the NTR facility. These scrams are not required and no credit for these scrams is used in the safety analyses. These scrams are not included in the Tech Specs. Nevertheless, GE has decided to conservatively maintain these scrams and they are included in the reactor standard operating and maintenance procedures. Additionally, there is a manual scram button in the reactor cell as shown in Section 7.3.2 and Figure 7-2 of the SAR.

Reference the response to General Question #1 for a description of other interlocks not mentioned in the SAR.

There are a number of alarms us:d for informational purposes which are not required by the Tech Specs. These are listed below with the SAR reference:

• Fuel loading tank high level, SAR 7.4 & 9.2 e hteactor cell door open. This is not discussed in the SAR.

• South cell door open, SAR 7.2

• Low core AP, SAR 7.4 e Low primary coolant flow. This is not discussed in the SAR.

• Cooling flow needed alarm. This alarm is activated when the reactor power exceeds 100 l watts. When this alarm activates, the low flow scram is armed. The low flow scram and bypass is discussed in SAR 7.3.2.

. Low count rate. This alarm is connected to the count rate recorder. This recorder traces the

source range monitor output. The alarm indicates that the source range monitor is i

abnormally low and may need to be adjusted for an on-scale reading. The source range I monitor is not in service and is not currently being used.

. Tech Spec Tables 3-1 and 3-2 provide all the required scrams and alarms.

Pane 7-5: This page indicates that bypasses are notprovided on most scram circuits. The

. nextpage discusses an automatic bypassfor lowprimary coolantflow while atpowers less l 'than 0.1 kW. Page 7-9provides a description ofother bypasses. Provide clanfication ofthe l bypassesfor reactor scrams, interlocks or other instrument and control systems and F components thatprovide safetyfunctions. Include when the bypasses are allowed. The suggested tablefor the preceding question may provide a convenient mechanismfor this l clarsfication.

Bypasses are not provided on most scram circuits. Scrams and interlocks are adequately l discussed in the response to General Question #1.

Pane 7-9: Provide venfication that thephysicalsystem does not allow bypass ofmore than one ofthepicoammeter trips. Provide renfication that when only two picoammeters are operable, that bypass ofa picoammeter is not allowed or describe the contrels or mitigation to allow such bypass.

l Administrative controls in the SOPS require that only one momentary bypass push button switch I be used at a time. Use of the bypass is permitted only during the moment that the range switch is

- moved to the next less sensitive range to prevent switching transients from causing a trip of the picolunmeter.

Pane 7-15: Provide a description of the leak testingprocedures and requirementsfor the neutron source. Include consideration offrequency and whether or not a Technical Specsfication requirement is needed.

(To be provided later.)

)

Pane 9-1: Reference to the ventilation system is to section 6.2. Section 6.2.1 does not a, ar in the section 6.2 on penetrations. Reference is also made to section 7.7for neutron source.

Provide clarification.

The correct references should be Sections 6.3 and 7.4.1 rather than 6.2 and 7.7.

Pane 9-1: Provide clan *fication on the statement that the NTR core is designed to last a infetime, since it is understood thatfuel burnup could result in the needfor additionalfuel by 2007.

The NTR core is commonly referred to as a lifetime core. Lifetime was considered to be a 1 nominal 40-year life. This would be either a 40-year license or a 20-year license with a 20-year license renewal. The NTR fue life is now estimated to be 50 years. Since the reactor has proven to be reliable and easily maintained, the life expectancy is now greater than initially expected.

Therefore, there is a conniet in common terminology versus reality. The older people retain the use of the term " lifetime core", while newer employees do not use this term. In fact, the NTR fuel, although not exceeding its own life, will apparently exceed the life ofits employees.

Panc 9-2: Provide verification that the reactor will be shut downforfires in the nuclear test reactor cell, and other specific locations. ,

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Section 4.3.2 of SOP 8.2, Non-reactor Emergencies, states,"A fire in the control room, the south cell or the adjacent areas (hallway, north room, shop) shall be cause for an immediate shutdown of the reactor." This SOP also states that if the Gre is not contained or is not small, shut down the reactor in accordance with procedures and evacuate. If there is a potential threat to personnel in the control room, the procedure requires an immediate reactor scram.

Pane 9-2: Provide clanfication of"small and contained"in the context offire definition.

"Small and contained" in the context of Hre dennition infers insignincant and not out of control, such as a trash can Dre, computer, copier or fax machine Sre or overheating. Locally mounted fire extinguishers provide rapid and adequate response. If the Dre cannot be controlled by the use of one fire extinguisher, it is no longer "small and contained". Building 105 is fully sprinklered.

Pnee 10-3: Provide verofication that there are procedural controls to enter the reactor cell during operations to ensure compliance to regulatory requirements.

Entry into the reactor cell is normally not peimitted when the reactor is critical. This has not I occurred in the last 20 years at least. Entry is permitted by SOP when a written procedure describing the work to be done is reviewed by the independent review body and approved by the fac ..y manager.

Pane 11-2: Provide a description ofor reference (e.g., section 16.1) to the other methodsfor detectingfuelleakage other than theprimary coolant sampling. Provide an analysis to demonstrate the sensitivities of these other methods to detectfuelfailure. This analysis should demonstrate that the combination of these other methods and the annualprimary coolant sampling are acceptable to detect and mitigate operation with leakingfuel to prevent unacceptable radiological exposure orfuel damage.

(To be provided later.)

Pane 12-3: Provide clarification that licensed Reactor Operators or licensed Senior Reactor

- Operators will direct the activities of trainees, and that licensed Senior Reactor Operators will

~ direct the activities oflicensed Reactor Operators.

l The activities of trainees, such as those described in 10CFR55.13 or 10CFR50.54(j), will be directed by a licensed reactor operator or licensed SRO. The licensed activities of operators such q as those described in 10CFR55.4, definition of senior operator, will be directed by a licensed senior operator.

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Pane 12-10: Provide or reference the definition of" abnormal occurrence"in section 12.3.2.

Provide or reference who can authorize restart after an abnormaloccurrence. Acceptable guidancefor LCO and Safety Limit violations and other events is contained in ANSUANS 15.1. This guidance is also applicable to section 12.4 on required actions. Additionally, all this information and guidance should satisfy the requirements of10 CFR 50.36(c)(1) which )

require NRC notification and restart authorizationfor safety limit violations. Time periods should alw be specipedfor reporting to NRC. The Technical Specifications (e.g., 6.5.2) shuuld also be verified to be consistent with this guidance.

Abnormal operation, as used in SOP 8.3, includes such things as the following:

e Unexplained reactivity changes while the reactor is critical e Power oscillations indicated on a neutron detection channel e Indication that movement of control or safety rods does not cause the expected reactivity change e Failure of any control or safety rod to respond normally to manual movement by the operator e High radioactivity indication on the stack monitor, continuous air monitor or remote area monitor There are other descriptions given in the SOP to describe abnormal operation. A definition was not formulated because an individual can identify abnormal behavior much better by learning a catalog of things rather than learning a definition which may not include all unforeseen circumstances. A normal situation is one that is easily recognized by a reasonably skilled and l i

experienced reactor operator. Conversely, the reactor operator is able to detect an abnormal situation, if the reactor were to be shut down because of an observed abnormal operation, the SOP requires a licensed SRO to be notified; the Manager, NTR, makes a determination if the abnormal operation is reportable and an SRO must approve restart of the reactor.

The SOP additionally contain a section which requires immediate shutdown for excessive reactivity or power changes and Tech Sepe violations. For Safety Limit violations, NRC approval would be obtained prior to resuming operation in accordance with Tech Spec 6.5.2.

l Pane 12-17: Provide verification in thefirstparagraph ofsection 12.9 that the QA program includes managerial and administrative controls to ensure safe operation in accordance as required. That is, QA is not limited to design and construction. Also, identify the differences between this QA program and the guidance in ANSI /ANS 15.8 " Quality Assurance Programs for Research Reactors. " Provide an analysis tojustify the differences or adopt the guidance of ANSI /ANS 15.8.

The NTR QA program in Section 12.9 is limited to design, fabrication and installation of hardware. Safe operation and maintenance of the reactor is assured by all the other programs listed in the SAR, such as Conduct of Operations, Radiation Protection Program, and Operator Requalification and are not specifically indexed in the QA program.

Pane 12-23: Provide clarification that the references arefor section 12 or 13. Ifnecessary provide corrections.

The references provided in Section 12.12 should have been in a separate section and are references for the entire SAR, not just for Section 12.

Pane 13-8: Provide clarification on the net reactivity characteristics that are said to be shown in Figure 13-2.

(To be provided later.)

Paec 13-9: Figure 13-2 seems to assume a pump startfrom 100 k W at time zero that is not allowed by Tech Specs. Provide additional explanation of Figure 13-2 to describe the transient or conditions it is depicting.

(To be provided later.)

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F .

l Pane 13-11: Figure 13-3 indicates "15-k WScram. " Provide clarification as the text and TS indicate 150 k Wfor scram. Venfy that this demonstrates that reactivity transientsfrom 150 k Wis more limiting thanfrom the low power conditionsfor allpotential reactivity transient conditions.

Figure 13-3 is mislabeled. It should read "Ak steps with 150-kW Scram" rather than "Ak steps l with 15-kW Scram".

Paec I3-30: Provide consistent terminologyfor " independent review unit" with that referred to in Section 12. Also, provide a more specific reference to the referred to " modulus operandi"in Section 12.

The " independent review unit" mentioned on page 13-30 is currently the group called

" Regulatory Compliance" which is discussed in Chapter 12. Their " modus operandi"is i delineated in Chapter 12 under the following headings:

l . 12.2.1, Composition and Qualifications

. 12.2.2, Charter

. 12.2.3, Review and Audit Functions Paec 13-31: Provide reference or clarificationfor the radiological consequence evaluation in section 13.5.3.1.

The radiological consequence evaluation in Section 13.5.3.1 refers to the radiological consequences of an accidental detonation of a device containing explosive materials discussed in Section 13.5.3. The evaluation was performed because these devices are subjected to the nondestructive test of neutron radiography. As a consequence, the devices are activated to low levels of short half-life isotopes.

Paec 13-35: Provide venfication ofproceduralcontrols within 50 feet ofexplosive devicesfor the operation of unshielded high-frequency generating equipment.

l This requirement is administratively controlled by Standard Operating Procedure 10.4, Explosive l Handling.

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Paec 13-37: Provide stack draft characteristics tofurther demonstrate the conservatism in the concentration calculation as indicated in thefootnote.

(To be provided later.)

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P m.

l Pane A-6: Provide clarification or reference (e.g., conduct ofoperations Chapter)for the 1 " Specialist,1ndustrialSafety and Hygiene" with regard to the organization location and responsibilities.

! On Figure 12-1, the Manager, V&MO, reports to upper level GE management. At the present time this is the President of GE Nuclear Energy. The Specialist, Industrial Safety and Hygiene,

( reports through other management levels to the President of GE Nuclear Energy. The office of

the Specialist, Industrial Safety and Hygiene, is located at the main San Jose site. The Specialist l works very closely with personnel at Vallecitos and regularly visits the site to attend VTSC (SAR Sections 12,12.2.2 and 12.2.3) meetings and to perform reviews and audits.

Pane A-9: Provide reference to the SAR Chapter and section that analyzed the expulsion of the beam preparation device.

l The reactivity effect of the expulsion of the beam preparation device is included in the 0.76$ step reactivity insertion analyzed in Section 13.4.3. Refer to Section 13.5.3.4.

l T3chnical Specifications

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( l Ppge 3-2: Provide or roference analyses that demonstrate the structure required to ensure a '

maximu.n U2 inch ma, ement on manualpoison sheets, and that demonstrate the reactivity effect of this allowed reactor core parameter.

l As discussed in the response to the question on SAR page 3-2, each Manual Poison Sheet (MPS) is latched in place to prevent movement. The Tech Spec limit is %-inch movement. The actual movement between the latch pin and the latch plate is a result of fabrication tolerances. The I i

measured movement is actually zero because the latch pin is spring loaded to provide a constant force. The Tech Spec limit of %-inch was chosen to allow for fabrication tolerances and also to l minimize the reactivity addition if the MPS were to move. The reactivity effect of a %-inch  !

j movement of the MPS would be extremely small (<5#). The MPS cadmium is 19 inches long and the fuel length is 15.25 inches long.

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Pnee 3-8: Provide clanfication on thefuelloading tank low water level alarm with regard to compliance to TechnicalSpecifications 3.3.3.1 and 3.3.3.2.

l Tech Spec 3.3.3.1 states that the reactor is cooled by light forced water coolant above 100 watts.

r Tech Spec 3.3.3.2 states that the core tank is filled with water during reactor operation. The fuel loading tank is connected to the reactor core tank as described in Sections 5.2,9.2 and Figures 1.1 and 5.1. When the water level in the fuel storage tank is above the low-level alarm point, it is assured that water is in the core tank and that forced water coolant is available when the primary cooling pump is operating.

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p-Teclinical Specifications (Continued)

Paec 3-13: Provide the basesfor TechnicalSpecification 3.4.3.4.

The stack monitoring equipment is required to be operating whenever activities are being conducted whica could release radioactivity. This assures that all radioactive releases are monitored prior to release to the environment.

Whenever the monitoring equipment is not operable, activities which could result in a radioactive release will be terminated.

l Pane 6-13: Currently Region IVdoes not have regulatory oversight responsibilityfor research reactors. Also, theprovisions of10 CFR 50.73(d) do not apply to research reactors (this l regulation is also mentioned on page 6-11). Current guidance in this regard is to send written j report to the U.S. Nuclear Regulatory Commission, A TTN: Document Control Desk, Washington D.C. 205S5. Provide clanfication.

Tech Spec 6.5.1.2 specifies that the occurrence "of the type identified in Section 6.5.2" be reported to the facility manager and reported to the NRC addressed in accordance with 10CFR50.73(d). Tech Spec 6.5.1.2 does not intend to refer to the nuclear power requirements i for reporting referred to in that Federal Regulation, only to the NRC address.

Tech Spec 6.6.1 requires that the annual report be submitted to the NRC Document Control Desk. A copy is sent to the Region IV Administrator. There are no regulatory requirements to send a copy to the P.egional Office. This is a courtesy copy only.

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