ML20196A863
| ML20196A863 | |
| Person / Time | |
|---|---|
| Site: | 07001113 |
| Issue date: | 05/16/1988 |
| From: | Winslow T GENERAL ELECTRIC CO. |
| To: | Rouse L NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| 24341, NUDOCS 8806300135 | |
| Download: ML20196A863 (64) | |
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Director Py/Sgf) h.
Office of Nuclear Material Safety and Sa s
U.S. Nuclear Regulatory Commission 9
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Washington, D.C.
20555 Attention:
Mr. L. C. Rouse, Chief
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7'o,i ici Fuel Cycle Safety Branch M/S WF1, 6-H-3
Dear Mr. Rouse:
Reference:
NRC License SNM-1097, Docket #70-1113
Subject:
License Amendment Request - Dry Conversion (Revision 21)
GE Nuclear Fuel and Components Manufacturing hereby submits a license amendment request to allow the conversion of UF, by the use of a new dry conversion process. is a description of the process and the safety systems. is a description of the requested amendment and several other minor license changes.
Attachr.ent 3 is the revised pages to edrt I of SNM-1097.
Our development program calls for use of enriched material during the third week of August, 1988.
We would appreciate your help in meeting this schedule.
Pursuant to 10 CFR 170.31, a GE check for $150 for processing this amendment request is enclosed.
If you have any questions regarding this matter, please contact me.
L].
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Vary truly yours,
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GE NUCLEAR ENERGY w:A y
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Preston Winslow, Manager f
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Mr.
L. C.
Rouse May 16, 1988 Page 1 of 20-ATTACHMENT 1 j
Objective The objective of utilizing dry conversion is to establish a less l
l complex process with reduced effluents, improved productivity and improved powder quality in our new product designs.
Successful dry process testing will provide an optional method of UF, conversion.
l Purpose The purpose of this submittal is to describe the dry process starting with UF, vaporization and ending at the point where material is introduced into the defluorinator.
At defluorination the remaining process is essentially the same as identified in our current NRC material license.
Criticality and radiological controls associated with the dry process are also described, including some of the beneficial reductions that are anticipated when compared with the existing ADU process.
Discussion The dry conversion process utilizes the existing UF, vaporization equipment to heat the UF, as feed material.
Once vaporized, the l
UF, gas flows via piping to a steam hydrolysis reactor where it is contacted with dry superheated steam to form anhydrous UO,F, powder and HF gas.
Separation of the powder and gas is performed using porous metal filters.
The UO,F, powder accumulates in a l
a l
Mr.
L. C.
Rouse May 16, 1988 1
Page 2 of 20 surge hopper located beneath the filter and the HF gas is drawn to a secondary filter under vacuum and then routed to the existing HF recovery process.
The UO F, is then fed to the defluorinator feed preparation unit to further prepare the powder for defluorination (Figure 1, Dry Conversion Block Flow Diagram).
Positive impacts from implementation of dry conversion, in coniparison to the existing ADU process, will include significant reductions in chemicals usage, reduced sludge generation for disposal, reduced 11guld waste generation, and less uranium sludge requiring recycle.
A decrease from the ADU process in actual environmental impacts is also expected via anticipated reductions as follows:
(a)
A 67% reduction of fluoride in the fluoride liquid waste stream as a result of steam hydrolysis and HF Recovery, and (b)
A resulting 67% reduction in calcium fluoride sludge for disposal since the recovered HF does not have to be treated.
It is anticipated that UO powder produced from this process will have improved ceramic characteristics such as powder flowability and pellet density characteristics for improved fuel quality.
Criticality and radiological safety requirements will be complied with as identified in US NRC Material License, SNM-1097.
9
. i Mr. L. C. Rouse May 16, 1988 Page 3 of 20 FIGURE 1 DRY CONVERSION BLOCK FLOW DIAGRAM 1
UF, VAPORIZATION 1
1 STEM 4 HYDROLYSIS REACTOR 1
1 e
SECONDARY PRIMARY (METAL) FILTER FILTER I
i 1
4 DEFLUORINATOR HF PREPARATION UNIT RECOVERY l
l A
1 4
TO EXISTING VACUUM DEFLUORINATOR SYSTEM
- ADDED EQUIPMENT i
i
Mr. L. C.
Rouse May 16, 1988 Page 4 of 20 Process 1)
UF, Vaporization (Existing)
UF, is received as a solid in 30-inch diameter cylinders.
The UF, is vaporized by heating the cylinder in a chamber through which air is circulated over electric heaters in the chamber.
After the UF, is vaporized, the gas exits the cylinder through the cylinder valve, passes through a flexible connection (pigtail), and into the UF, piping.
To prevent solidification of the UF, gas, all piping and valves are heated.
The UF, pressure from the cylinder is controlled via the heater.
The.
skin temperature of the cylinder and the temperature inside the air recirculation duct are also monitored to-prevent overheating.
If the cylinder temperature or the pressure reaches preset limits, the heaters automatically switch off.
Some of the safety features of the chambers include emergency valve shut-off handles located outside the vaporization room and photo eyes that detect gas leaks in the chamber and alarm in the control room for investigation.
Purging capabilities exist for all the UF, lines and is a feature between vaporization and the steam hydrolysis
- reactor, j
1 The existing vaporization room is designed to contain UF, releases and prevent its escape to the environment.
Mr.
L. C. Rouse May 16, 1988 Page 5 of 20 2)
Steam Hydrolysis Reactor (Added Equipment)
The UF, gas flows from the heated UF, cylinder to the steam hydrolysis reactor via heated piping.
This maintains the UF, in gaseous form to assure an even flow and prevent in-line i
condensing.
When the pre-determined process parameters are verified, the UF, is allowed to flow into the steam hydrolysis reactor.
Once in the reactor, the UF, gas is contacted with dry superheated steam in a safe geometry pipe vessel to form anhydrous UO F powder and HF gas.
The reaction between UF, and the steam is instantaneous and exothermic.
The spontaneity of the reaction assures essentially immediate and complete conversion to solid UO F, particles.
Heat from the reaction will help keep the gases at a temperature above the dew point of the steam and prevent reactor pluggage, s
Pressurization is prevented from developing in the reactor by maintaining and monitoring the reactor under a negative pressure.
If a parameter in a particular process section goes out of specification, a computer control system will shut off the UF, flow which in turn will shut down that part of the i
- process, Free hydrogen is not generated in the hydrolysis chemical reaction.
Hydrogen from dissociated mEmonia (DA) is currently present downstream at the defluorinator.
Backflow from the defluorinator is prevented using nitrogen purged air locks.
The defluorinator ventilation system is not connected to the steam hydrolysis ventilation system.
Mr.
L. C.
Rouse May 16, 1988 Page 6 of 20 3)
Filter, Feeder, and Defluorinator Feed Preparation Unit (Added Equipment)
Seoaration of UO,F, and HF gas is performed in a porous metal primary filter.
The UO F, powder accumulates in an 8" diameter pipe surge hopper located beneath the filter and the HF gas is piped to a secondary filter and then to an existing HF recovery unit.
The UO,F, powder is fed from the pipe hopper through triple rotary airlocks to the defluorinator feed preparation unit from which the treated material passes directly to the screw feeder on the front end of the existing defluorinator.
Pressure, temperature and flow from vaporization to the defluorinator feed point are monitored and controlled.
Exceeding a safety related parameter will cause the system to react in a manner to mitigate any unsafe conditions.
This reaction will alarm and, if necessary, shut down the system.
4)
HF Recovery (Existing)
Hydrofluoric acid gas and excess steam are removed from the steam hydrolysis offgas in a favorable geometry HF recovery 4
system after the gases pass through a secondary filter which is a contingency to the primary filter.
Most of the HF and steam is condensed as the gas stream passes through an offgas cooler.
The residual HF and steam is recovered by countercurrent scrubbing of the gas stream with deionized water in an HF stripping column.
The recovered HF is then purified by rebo111ng the bottoms of the stripping column and
Mr. L. C. Rouse a
May 16, 1988 Page 7 of 20 condensing the vapor overheads as by-product HF.
This vaporization step assures that uranium scrubbod from the offgas stream will not contaminate the by-product HF beyond trace quantities.
For employee protection the HF recovery system is enclosed in an isolated room which is conctructed of HF resistant material.
Entry is not permitted when the system is operating.
Thus, the liquid HF is isolated.
The offgas system is under atrong vacuum and leaks are inward during operation.
Excessive leaks cause loss of vacuum which shuts the process down.
The transfer of liquid HF from the collection tank to storage is through fluorocarbon lined steel pipe which has. flange cuards.
5)
Vacuum System (Existing) l Vacuum fer the dry conversion process is generated in a multiple eductor system where fluoride waste from the discharge of uranium recovery is pumped at high volume through the eductors and recirculated through two rundown tanks.
The fluoride waste contains ammonium hydroxide which neutralizes any HF remaining in the offgas.
The rundown tanks are vented to the main building HVAC system prior to the final exhaust scrubber and HEPA filtration.
This assures HEPA filtration of the 'temaining dry conversion process offgas before discharge.
l Mr.
L. C.
Rouse May 16, 1988 Fege 8 of 20 Safety 1)
Criticality Safety The dry process utilizes existing, approved equipment, i.e.
vaporizers, HF recovery and defluorinator, as well as the described new equipment.
The new equipment (steam hydrolysis
+
reactor, primary filter, rotary air locks, and defluorinator feed preparation unit) are analyzed as safe geometry using techniques and limits defined in SNM-1097, Chapter 4.
The steam hydrolysis reactor rotary air locks and defluorinator feed preparation unit each have a cylindrical shape of less than 9" inside diameter.
This is a safe diameter based on Table 4.1 of license SNM-1097.
The primary filter is a 16" outer diameter (OD) cylindrical vessel containing numerous 2" OD porous metal filter elements.
This filter was analyzed by the KENO code and determined to have a neutron multiplication i
j factor of less than.90 under normal conditions and less than
.97 under accident conditions.
Engineered features in the equipment design will preclude the possibility of criticality should favorable geometry control be lost, satisfying the j
double contingency principle, 2)
Radiological Safety Radiological safety considerations relate primarily to airborne hazards.
The system is operated with a negative pressure and ventilated hoods are provided at normal
)
access / maintenance points.
Bioassay and air sampling programs will be administered according to our current programs.
)
i
4 Mr. L. C. Rouse May 16, 1988 Page 9 of 20 l
i 3)
Backflow Control i
The backflow of material into unsafe and/or unintended i
equipment will be prevented using a control system that j
monitors key safety aspects of the process.
The control system and safety related valves are designed to actuate to i
safe shutdown conditions in the event of loss of instrument air, nitrogen, or electrical power.
l 4
The primary backflow control of gases and powder in the steam I
hydrolysis reactor relies upon.ngative pressure used to l
transport material from the UF, cylinder towards the reactor and then towards the primary filter.
The control system monitors the negative pressure and shuts off UF.,
steam and j
air / nitrogen flows using automatic block valves should negative pressure be lost.
t In addition to negative pressure, temperature and the flow of i
gases are monitored and controlled with high and/or low 1
setpoints to ensure safe and efficient transfer.
The backflow of combustible gases from the defluorinator operation are prevented by segregation of the defluorinator l
from the process by use of nitrogen-purged rotary airlocks and a separate defluorinator ventilation system.
j I
The control system has pre-set parameters (high and/or low) j for key components of the process.
Unless key safety parameters are satisfied, the process will not generate or restart after a perturbation.
l l
Mr. L. C. Rouse May 16, 1988 Page 10 of 20 4)
_ Safety Interlocks A systemctic hazards analysis approach is used during process design to identify safety controls which mindsdze potential criticality, radiological, and chemical hazards.
Selection of criticality safety controls is based on the double contingency principle.
There are a number of elements which are considered when establishing safety controls.
Examples include equipment geometry, process containment, structural integrity, safety interlocks, material of construction, operating procedures, moderator content, common mode failure, and credible accident conditions.
While this list is not all-inclusive, it shows that process interlocks and alarms are an integral part of. the overall safety controls.
Interlocks apply automatic actions to prevent startup and/or to shut down the process or part of a process before potentially unsafe conditions develop.
Alarms indicate potentially hazardous process conditions to the operator.
For the dry conversion process, interlocks are primarily used to satisfy the second criticality safety contingency and to prevent backflow.
Figure 2 shows the principal interlocks incorporated into the dry conversion process design.
These interlocks are:
a)
Steam hydrolysis reactor pressure b)
Steam to UF, flow ratio c)
Superheated steam temperature d)
Steam hydrolysis reactor temperature e)
Primary filter temperature f)
Secondary filter differential pressure 1
9 Mr.
L. C.
Rouse l
May 16, 1988 Page 11 of 20 g)
HP stripping column level h)
HP stripping column D.I. Water flow i)
Defluorinator feed preparation pressure A description of each interlock follows:
a)
Steam Hydrolysis Reactor Pressure Interlocks _
Backflow prevention and containment of uranium powdera and hydrofluoric acid is assured in the dry conversion process by operating the process from steam hydrolysis through hydrofluoric acid recovery under negative pressure.
Two reactor pressure controls independently shut off UF.,
steam, dry air, and N, purge feeds to the reactor'if either or both instruments detect loss of vacuum.
The pressure controls are diverse to the JXtent that one is analog whereas the other is a high pressure switch.
The analog pressure controller also controls the main vacuum control valve.
The failure state for the main vacuum control valve is in the open position.
The failure state for all feed shutoff valves is the closed position.
These two pressure interlocks prevent backflow of UF, or uranium powder into the steam, dry air, and nitrogen supplies as well as the backflow of steam into a UF.
cylinder.
A6ditionally, vacuum at the reactor, which is the most distant piece of equipment from the vacuum source, assures uranium and hydrofluoric acid containment in the steam hydrolysis reactor, primary filter, secondary i
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ALARM FIGURE 2 PRINCIPAL DRY CONVERSION INTERLOCKS AND ALARMS 9
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Mr.
L. C.
Rouse May 16, 1988 Page 13 of 20 filter, offgas condenser, HF stripping column and i
interconnecting piping, j
l The two independent interlocks satisfy double antingency for backflow prevention.
b)
Steam to UF, Flow Ratio Interlock t
Complete conversion of uranium hexafluoride (UF.) to uranyl fluoride (UO,F,) is assured when the mole ratio of steam to UF, is greater than the stoichiometric l
requirement of 2.
t UF,
+.2,H,0 - UO,F, + 4 HF i
The control system continuously calculates the ratio of steam mass flow to UF, mass flow.
If the ratio is less I
than 1.25 time the stoichiometric requirement of 2, the control system shuts off the UF, flow.
The factor of 1.25 provides a 25% safety margin.
The failure state of the UF, valve is the closed position.
If the flow ratio interlock fails and insufficient steam flow is available to convert all the UF., UF, will flow to the HF recovery system where the UF, will condense in the offgas condenser and/or,be hydrolyzed 2nd scrubbed in the l
HF stripping column.
Both pieces of equipment are of favorable geometry.
The hydrolyzed UF, will be trapped in the bottom rebo11er portion of the stripping column.
i Periodically, the stripping column bottoms are discharged l
n-y
Mr.
L. C.
Rouse May 16, 1988 Page 14 of 20 to the safe geometry fluoride waste quarantine system to assure purging of any uranium in the unlikely event this incident occurs.
Uranium discharged to the quarantine system is precipitated with ammonium hydroxide and removed by the existing fluoride waste inertial filter and centrifuge system.
Prior to release to uranium recovery (UPMP), the fluoride waste stream uranium concentration is measured to assure acceptable discharge values.
c)
Superheated Steam Temperature Interlock, Although the dry conversion process uses caly favorable geometry, a second control is needed to meet the double contingency principle when a catastrophic failure of geometry scenario is postulated.
The reactor pressure interlocks assure shutdown of the process, but the equipment could possibly contain greater than a critical mass of uranyl fluoride.
Criticality safety is maintained by operating at conditions which produce anhydrous uranyl fluoride.
If reaction temperature and equipment temperatures are greater than the uranyl fluoride dehydration temperature, the uranyl fluoride will be anhydrous 2 with a moisture content less than 1.0 weight 1
percent.
This provides moderation control as a secondary contingency.
Anhydrous uranyl fluorido also minimizes the risk of equipment plugging.
Therefore, the superheated steam supply temperature is maintained to assure the production of anhydrous uranyl fluoride and the avoidance of steam condensate.
1Katz, J.F.
and Rabinowitch, E.,
The Chemistry of Uranium, McGraw-Hill Book Company, New York (1951),
p.
559 9
Mr. L.
C. Rouse May 16, 1988 Page 15 of 20 j
Two trip limits are programmed for the superheated steam interlock.
Below the first or low trip limit, the UF, feed is shut off.
If the temperature drops further to the low low trip limit, the steam supply is shut off to avoid condensing steam.
For both the UF, feed valve and the steam supply valve, the failure state is the closed position.
d)
Steam Hydrolysis Reactor Temoerature Interlock The steam hydrolysis reactor temperature interlock is similar to the superheaced steam temperature interlock.
The only difference is that two reactor temperature detectors are used.
If either or bo h instruments detects low temperature conditions, the interlock control will shut off UF, flow and/or steam flow to the reactor.
A two tier trip point is used with the same trip temperatures as the superheated steam interlock.
Likewise the low temperature trip limit shuts off the UF, supply and the low low temperature trip limit shuts off the steam
- supply, e)
Primary Filter Temperature Interlock The primary filter temperature interlock is identical to the steam hydrolysis reactor temperature interlock, f)
Secondary Filter Differential Pressure Interlock The secondary filter is a contingency to prevent uranyl fluoride from entering HF recovery if a primary filter
1
\\
l Mr. L.
C.
Rouse May 16, 1988 Page 16 of 20 porous metal element should fail.
The secondary filter has both a low and a high differential pressure interlock.
High differential pressure may be an indicator of primary filter tube failure in which case the secondary filter will retain the uranyl fluoride.
Low secondary filter differential pressure may indicate that the secondary filter has been assembled improperly or that the filter has ruptured.
The high and low differential pressure interlocks shut down the steam hydrolysis reactor, the primary filter and the dry air purges.
The steam hydrolysis reactor shutdown closes the UF.,
steam, dry air, and N, feed valves.
The primary filter shutdown closes the N, blowback and the dry air purge valves.
The dry air purges to the secondary filter differential pressure sensor are shut off to prevent uranium backflow.
The failure state for all valves is the closed position.
g)
HP Stripping Column Level Interlock A single contingency prevents UF, (steam to UF, flow ratio interlock) or uranyl fluoride (secondary filter differential pressure interlock) from entering the HF recovery system.
Although the HP recovery system is constructed of favorable geometry equipment, an overflow of the system in-conjunction with a failure of the first contingency, could potentially flow uranium to the non-favorable geometry vacuum system.
Without an overflow, only HF and water vapors could potentially enter l
I
'l Mr. L.
C.; Rouse May 16, 1988 Page 17 of 20 i
-the vacuum system.
Uranium would be trapped in the bottom j
reboiler portion of the stripping column.
Therefore the HF stripping column level is a process interlock.
If the stripping column level is high, the.UF, feed and steam supply to the steam hydrolysis reactor and the deicnized water supply to the stripping column are j
shut'off.
Also the stripping column rebo11er steam supply i
is shut off to prevent bubble lifting of solution or j
carryover into the HF collection tank..The closure of the l
deionized water suroly valve additionally serves to i
prevent uranium backflow into the deionized water system.
The failure state for all valves is the closed position, f
h)
HP Stripping Column D.I. Water Flow Interlock f
If UF, should enter HF recovery after a contingency
'j failure, the UF, would be hydrolyzed to uranyl' fluoride j
and scrubbed from the offgas stream in the HF. stripping j
column.
Likewise, if uranyl fluoride powder should enter l
HF recovery af ter a contingency f af. lure, the uranyl i
fluoride will be scrubbed from the offgas stream in the HF stripping column.
But this scrubbing action will only occur if water is supplied to the column.
l The deionized water flow to the HF stripping column is therefore a process interlock.
If the flow should drop below the low flow trip point, the UF, feed to the steam hydrolysis reactor is shut o'ff.
The deionized water flow control valve failure state is in the open position.
The
l'
)
Mr.
L. C.
Rouse May 16, 1988 Page 18 of 20 flow control valve is separate from the valve that is closed by high level in the HF stripping column.
The combination of HF stripping column high level interlock and HF stripping column low D.I. water flow interlock satisfy the second contingency to prevent uranium from entering the non-favorable geometry vacuum system.
1)
Defluorinator Feed Preparation Pressure Interlock Nitrogen purges are used in conjunction with three rotary airlocks in series to separate the strong vacuum of the primary filter from the defluorinator ventilation system.
To avoid hydrogen backflow from the defluorinator, the nitrogen purge area below the retary airlocks must be at a pressure greater than the defluorinator.
This is accomplished by a pressure dc*.ector'above the defluorinator feed preparation unit which is interlocked to the defluorinator hydrogen supply..
If the pressure is below a low trip point, the defluorinator hydrogen is shut off.
This assures a flow of nitrogen from the rotary airlocks through the feed preparation unit to the defluorinator without hydrogen backflow.
The failure state for the defluorinator hydrogen valve is the closed position.
5)
Conditions for Process Operation i
The computer control system for the dry conversion process has a prestartup sequence which assures the. principal safety
Mr.
L. C.
Rouse May 16, 1988 Page 19 of 20 interlocks and controls are functioning prior to startup with UF.
This startup sequence in order by process unit is:
(1) Vacuum system (2) HF recovery (3) Primary and secondary filters and rotary airlocks (4) Steam hydrolysis reactor (5) Defluorinator feed preparation unit (6) UF, flow UF, vaporization and defluorination is independent of the startup sequence but must be operational prior to UF, flow being initiated.
Each unit of the process must clear its safety interlocks before the startup can proceed to the next unit.
The UF, flow section is a master program which monitors each process unit for interlock clear status and the principal safety interlocks.
Operating parameters for the dry conversion process are either programmed into the computer or are set by operating procedures.
In either case the operating parameters for safety interlock controls shall be conservative with respect to the limiting interlock trip values.
Emergency shutdown of the process or appropriate portion of the process is automatic if a safety interlock trips.
The a
v
Mr.
L.
C. Rouse May 16, 1988 Page 20 of 20 control room operator also can shut off UF, flow with an emergency switch..
Furthermore,' operators can manually shut off any flow if the automatic valves fail to function.
Loss of power to the control system or any device will cause the process or appropriate port' ion of the process to fail to-a safe state.
All valves are of a type that fail closed except the vacuum system valves and the HF stripping column D.I.
water flow control valve which are of a type that fail open.
6)
Management Controls Preventive mainteaance, calibration and periodic function tests assure the process safety interlocks remain reliable.
Preventive maintenance and calibration is part of the automated maintenance system.
For the dry conversion process principal safety interlocks, calibration shall be carformed at least semi-annually during line operations.
If the orocess has been shut down and more than six months.have elapsed since the previous calibration, the principal-safety interlocks shall be calibrated prior to startup.
Functional testing of the principal safety interlocks from the sensor to the final control element shall be performed at least annually after startup.
If the process has been shut down and more than one year h'as elapsed since the previous functional test, each principal safety interlock shall be functionally tested prior to startup.
Mr.
L. C.
Rouse May 16, 1988 Page 1 of 1 P
ATTACHMENT 2 DESCRIPTION OF REQUESTED REVISIONS Page Section Description f
I-1.4 1.4.4 Change of wording from "materials" to "rods" as previously stated in the license.
I-1.6 1.7.1.1 Added new wording to allow for the I-1.8 and conversion of UF, using dry conversion as 1.7.2.7 described in NF&CM's application dated 5/16/88.
I-1.10 1.8.2 Clarification of the section's intent by describing the exemption and correcting erroneously stated page numbers.
I-2.8 2.3.2 Revised to clarify the frequency of the Operational Radiation Safety Committee's meeting.
I-2.15 2.7.1 Reference to the applicable footnote (2) has been added to the word "Annually".
This reference had been unintentionally omitted.
I-5.6 5.1.1.4 Change to allow a different type of counter for the stack sample filters.
I-6.3 6.5 Added a new section describing the safety I-6.4 interlock controls used in the dry I-6.5 conversion (steam hydrolysis) process.
I-6.6
}
Mr.'L. C. Rouse May 16, 1988 Page 1 of 1 ATTACHMENT 3 NRC License SNM-1097 page changes to Part I.
Chapter 1 has been provided~as_a total replacement because some of the changes affected page numbering.
_ Changes to Chapters 2 and 5 are presented as individual page changes.
All changes are indicated with an asterisk (*) to the right of the change, i
REVISIONS BY PAGE Effective Effective Effective Page Date Page Date Page Da,te I-3.6 10/23/87 l
TABLE OF CONTENTS I-1.21 5/16/88 I-3.'
I-1.22 1
10/23/87 I-1.23 I-3.8 I-3.9 2
I-1.24 3
I-3.10 j
4 I-3.11 5
CHAPTER 2 I-3.12 6
I-3.13 7
I-2.1 10/23/87 I-3.14 8
5/16/88 I-2.2 I-3.15 I-2.3 I-3.16 9
I-2.4 I-3.17 10 I-2.5 I-3.18 11 12 I-2.6 I-3.19 13 I-2.7 I-3.20 I-2.8 5/16/88 I-3.21 I-2.9 10/23/87 I-3.22 PART I I-2.10 I-3.23 I-2.11 I-3.24 I-2.12 I-3.25 CHAPTER 1 I-2.13 I-3'.26 I-2.14 I-3.27 I-1.1 5/16/88 I-2.15 5/16/88 I-3.28 I-1.2 I-2.16 10/23/87 I-1.3 I-2.17 I-1.4 I-2.18 I-1.5 I-2.19 CHAPTER 4 I-1.6 I-2.20 I-1.7 I-2.21 I-4.1 10/23/87 I-1.8 I-2.22 I-4.2 I-1.9 I-2.23 I-4.3 I-1.10 I-2.24 I-4.4 I-1.11 I-4.5 I-1.12 I-4.6 I-1.13 I-4.7 I-1.14 CHAPTER 3 I-4.8 I-1.15 I-4.9 I-1.16 I-3.1 10/23/37 I-4.10 I-1.17 I-3.2 I-4.11 I-1.18 I-3.3 I-4.12 I-1.19 I-3.4 I-4.13 I-1.20 I-3.5 I-4.14 LICENSE
$NM-1097 OATE 5/16/88 Page DOCKET 70-1113 REVISION 21 L
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i LICENSE SNN-1097 DATE 5/16/88 Page DOCKET 70-1113 REVISION 21 -
r CHAPTER 1 STANDARD CONDITIONS AND SPECIAL AUTHORIZATIONS 1.1 CORPORATE & FINANCIAL INFORMATION This licensing information document is filed by the Nuclear Fuel & Componont Manufacturing facility (identified in this document as GE-Wilmington) of the General Electric Company, a New York corporation with the principal place of business at Schenectady, New a
York.
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1.2 LOCATION & GENERAL DESCRIPTION OF WILMINGTON PLANT The GE Nuclear Fuel & Components Man"*.acturing operates a nuclear fuel fabrication plant in Wilmington, North Carolina.
At this site, GE occupies buildings for administrative, laboratory and manufacturing activities.
A site plan is included as Figure 1.1.
Fuel manufacturing activities are conducted within the fuel manufacturing area.
The full address is as follows:
GE Nuclear Energy.
Nuclear Fuel & Components Manufacturing, (name of person and mail code),
P. O.
Box 780, Wilmington, NC 28402.
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1.3 LICENSE NUMBER The GE-Wilmington NRC license number is SNM-1097 (Docket 70-1113).
In accordance with the GE-Wilmington timely renewal
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request dated 4/28/81 and subsequent, related submittals, GE-Wilmington requested license renewal for a five year period commencing from the time that the Nuclear Regulatory Commission completed final action on that matter.
1.4 POSSESSION LIMITS The following types, quantities, and forms of special nuclear materials are authorized:
1.4.1 Uranium-235, 50,000 kas total Contained in uranium to a maximum, nominal enrichment of 6% in the form of UF, UO,, U 0, and other solid and 3
liquid process intermediates and products characteristic of LEU fuel fabrication and fuel fabrication development activities.
1.4.2 Uranium-235, 350 grams total, In any form contained in uranium at any enrichment, for use in measurements, detection, research or development.
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1.4.3 Uranium-235, 500 kas Total In any form, contained in uranium at a maximum nominal enrichment of 15% for use in laboratory and process technology development operations.
1.4.4 Plutonium 1 milligram in samples for analytical purposes 1 milligram as standards for checking the alpha radiation response of radiation detection instrumentation.
20 grams as sealed neutron sources In nuclear fuel rods at a level of not more than 10-5 grams of plutonium per gram of U885 1.5 MATERIAL USE LOCATIONS Uranium normally will be used at the Wilmington site in the fuel manufacturing area only.
Conversion and fabrication of SNM is conducted within the fuel manufacturing building.
Small qua'ntities (i.e.,
less than one safe batch of uranium in a non-dispersable form) may be temporarily moved to other buildings or site locations for special tests.
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1.6 DEFINITIONS Definitions as used in the license conditions are set forth below.
1.6.1 Area Manager - The manager designated by the GE-Wilmington facility manager who is responsible for implementetion of nuclear safety requirements in the area assigned.
The general title "Area Manager" does not necessarily refer to the title of any specific position in the GE system of organization and position nomenclature.
1.6.2 Array - means two or more interacting accumulations of fissile material.
1.6.3 Criticality control - means the administrative and technical requirements established to minimize the possibility of achieving inadvertent criticality in the environment analyzed.
1.6.4 Full Reflection - means the degree of reflection equivalent to a tight fitting shell of 12 inches or more of water.
1.6.5 Minimal Reflection - means the degree of reflection equivalent to a close-fitting shell of water, steel, aluminum, nickel or copper not greater than 1/8 inch in thickness.
1.6.6 Minimum Critical Dimension - means the smallest dimension which constitutes a critical system for a i
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e given geometry under conditions of full reflection and optimum moderation.
1.6.7 Nominal U ss Enrichment - means the value of enrichment assigned to a quantity of material for convenience because the precise value is not known due to process or-measurement uncertainties.
1.6.8 Nuclear Safety - means both criticality and radiation safety.
1.6.9 Safe Batch - means an accumulation of special nuclear material which is 45% of the critical accumulation for U285 enrichments less than or equal to 6% and 33 1/3% of the critical accumulation for U885 enrichments greater chan 6%, considering enrichment, full reflection, and optimum water moderation for the specific material form.
1.7 AUTHORIZED ACTIVITIES This application requests authorization to receive, possess, use, store and ship duthorized special nuclear materials pursuant to 10 CFR Parts 70, 71, 73, 74 and 75.
1.7.1 Product Processing Operations 1.7.1.1 UF, Conversion - Conversion of uranium hexafluoride to uranium oxides by ADU, GECO, and a dry conversion process.
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1.7.1.2 Fuel Manufacture - Fabrication of nuclear reactor fuels containing uranium.
1.7.1.3 Scrap Recovery -
Reprocessing of unirradiated scrap from GE-Wilmington and from other sources with nuclear safety characteristics similar to GE-Wilmington in-process materials.
I 1.7.2 Process Technology Operations 1.7.2.1 Devclopment and fabrication of reactor fuel, fuel elements a'nd fuel assemblies in small amounts or of advanced design.
1.7.2.2 Development of scrap recovery processes 1.7.2.3 Determinaticn of interaction between fuel additives and fuel materials.
1.7.2.4 Chemical analysis and material testing, including physical and chemical testing and analysis, metallurgical examination and radiography of uranium compounds, alloys and mixtures.
1.7.2.5 Instrument research and calibration, including development, calibration and functional testing of nuclear instrumentation and measuring devices.
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1.7.2.6 Other process technology development activities related.
to, but not limited by, the above.
1.7.2.7 Conversion of UF, to UO and other intermediate compounds using a dry process.
1.7.3 Laboratory operations Chemical, physical or metallurgical analysis and testing of uranium compounds and mixtures, including but not limited to, preparation of laboratory standards.
1.7.4 General Services Operations 1.7.4.1 Storage of unirradiated fuel asttmblies, uranium compounds and mixtures in areas arranged specifically for maintenance of criticality and radiological safety.
1.7.4.2 Design, fabrication and testing of uranium prototype processing equipment.
1.7.4.3 Maintenance and repair of uranium processing equipment and auxiliary systems.
1.7.4.4 Storage and nondestructive testing of fuel rods containing licensed amounts of plutonium.
1.7.5 Waste Treatment and Disposal 1.7.5.1 Treatment, storage and disposal and/or shipment of liquid and solid wastes whose discharges are regulated.
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6 1.7.5.2 Decontamination of non-combustible contaminated wastes to reduce uranium contamination levels, and subsequent shipment of such low-level radioactive wastes to j
licensed burial sites for disposal.
1.7.5.3 Treatment or disposal of combustible waste and scrap material by incineretion pursuant to 10 CFR 20.302(a) and 10 CFR 20.305.
1.7.6 Offsite Activities Testing, demonstration, non-destructive modification, and storage of r.aterials and devices containing unirradiated uranium, provided that such materials and devices shall be in GE control at all times.
1.8 EXEMPTIONS & SPECIAL AUTHORIZATIONS 1.8.1 Requirements for Prior Authorization of Activities by License Amendment Prior authorization by license amendment shall be required for the following activities:
1.8.1.1 Major changes or additions to existing processes which may involve a significant increase in potential or actual environmental impact resulting from utilizing such changes or additions.
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1.8.1.2 Major process changes or additions which involve a new process technology for which a criticality safety demonstration has not been previously submitted to the Commission.
In determining whether a new process technology requires such prior authorization by license amendment, the following factors will be considered:
(1) type of equipment utilized, (2) chemical reactions involved and (3) potential and/or actual environmental impact.
1.8.1.3 Proposed activities for which specific application and prior approval are required by Commission regulations..
1.8.2
. Contamination-Free Articles Authorization to use the guidelines, contamination and exposure rate limits specified in license pages I-1.20 through I-1.23, "Guidelines for Decontr.mination of Facilities & Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Eyproduct, Source, or Special Nuclear Material," US NRC, August 1987 for decontamination and survey of surfaces or premises and equipment prior to abdandonment or release for unrestricted use.
1.8.3 Disposal of Contamination-Free Liquids 1.8.3.1 Hydrogen Fluoride Solutions Authorization, pursuant to 10 CFR 70.42(b) (3), to transfe liquid hydrofluoric acid to Brush Wellman, i
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Elmore, Ohio, through the chemical supplier, Consolidated Chemical Company, Kansas City, Missouri, without eitl'er company possescing a NRC or Agreement State license for special nuclear material, provided that the concentration of uranium does not exceed three parts per million by weight of the liquid and the nominal enrichment does not exceed 6 weight percent of U2'5 The hydroflucric acid is transferred and used in such a manner that the minute quantity of uranium does not enter into any food, beverage, annmetic, drug or othar commodity designated for ingestion or inhalation by, or application to, a human being such that the uranium concentration in these items would exceed that which naturally exists.
Additionally, the acid is used in a process which will not release the low levels of radioactivity to the atmosphere as airborne material and whose residues will remain in a lagoon system.
Prior to shipment, each transfer is sampled and measured to assure that tne concentration does not exceed three parts per million of uranium.
GE shall maintain records under this condition of license including, as a minimum, the date, uranium concentration and quantity of all hydrofluoric acid transferred.
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1.8.3.2 Nitrate-Bearing Liquids
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Authorization, pursuant to 10 CFR 20.302(a), to dispose of nitrate-bearing liquids, provided that the uranium concentration does not exceed a 30-day average of 5 parts per million by weight of the liquids and the nominal enrichment does not exceed 6 weight percent U885, by transport to an offsite liquid treatment system located at Federal Paper Board Corporation, Riegelwood, North Carolina, in which deccmposition of~the nitrates will occur and from which the denitrified liquids will be discharged in the effluent from the system.
1 The environmental monitoring program as described in Section 5.1.4.2 is used to control these activities.
1.8.4 Use of Materials at Off-Site Locucions 1.8.4.1 Authorization to use up co 15 grams of U225 at other sites within the limits of the United Staces and at temporary job sites of the licensee anywhere in the United States where the Nuclear Regulatory Commission maintains jurisdiction for regulating the use of licensed material.
The manager of the radiation safety function shall establish the safety criteria for material being used at offsite locations.
It is also his responsibility to designate the individual who will be responsible to carry out these criteria.
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1.8.4.2 Authorization to store at nuclear reactor sites, uranium fully packaged as for transport in any Fissile Class I package, in accordance with the conditions of a icense authorizing delivery of such containers to a carrier for Fissile Class I transport, at locationt in the United States providing such locations are controlled by GE with provision to minimize the severity of potential accident conditiens to be no greater than these in the design bases for the containers during transportation.
Provisions for compliance with applicable 10 CFR 73 requirements are described in the NRC-approved GE-Wilmington Physical Security Plan dated June 6,
- 1986, as currently revised in accordance with regulatory provisions.
Storage at nuclear reactor sites is subject to the financial protection and indemnity provision of 10 CFR 140 and is limited to possession for purposes of delivery to a carrier for transport.
The requirements of 10 CFR 70.24 are waived insofar as this section applies to the materials contained in Fissile Class I packages.
1.8.4.3 Authorization to store at nucl. ~? reactor sites, arrays of finished I; actor fuel rods and/or assemblies in any of the inner metal containers of the RA-series shipping package described in package certificate USA /4986/AF, 6t locations in the United States providing such locations are controlled by GE with provision to minimize the severity of potential accident conditions to be no l
LICEt!SE SNM-1097 OATE 5/16/88 Page DOCKET 70-1113 REVISION 21 I-1.13
greater than those in the design bases for the containers during transportation.
Arrays can be constructed without limit to the number of containers so stored, except that each array shall be stacked to a height of no more than 4 containers high with each container separated by nominal 2 inch wooden studs, and with the width and length for each-array and separation between arrays determined only by container handling requirements.
Provisions for compliance with applicable 10 CFR 73 requirements are described in the NRC-apprcved GE-Wilmington Physical Security Plan dated June 6, 1986, as currently revised in accordance with regulatory provisions.
Storage at nuclear reactor sites is subject to the financial protection and indemnity provision of 10 CFR 140 and is limited to possession for purposes of delivery to a carrier for transport.
The requirements of 10 CFR.70.24 are waived insofar as this section dpplies to the materials contained in any of the inner metal containers of the RA-series shipping package.
(Reference Section 1.8.7).
1.8.4.4 Authorization to transfer, possess, use and store unirradisted reactor fuel )f GE manufacture at nuclear reactor sites, for purposes of inspection, fuel bundle disassembly and assembly, including fuel rod LICENSE SNN-1097 DATE 5/16/88 Page-DOCKET 70-1113 REVISION 21 I-1.14 l
replacement, provided that the following conditions are met.
1.8.4.4.1 A valid NRC license has been issued to the reactor licensee, which authorizes receipt, possession and storage of the fuel at the reactor site, and that GE possesses the fuel only within the indemnified location.
1.8.4.4.2 Not more than one fuel assembly and 30 unassembled fuel rods of the types described in NRC Certificate of Compliance USA /4986/AF, are possessed by GE at any one reactor site at any one time, except when the fael has been packaged for transport.
1.8.4.4.3 All,,erations involving the fuel are conducted by or under the direct supervision of a member of the GE staff who shall be responsible for all work on the fuel element assembly.
The person shall be knowledgeable and shall have access to all_ applicable procedares and license conditions at the reactor site and the appropriate actions that are to be taken in the event of emergencies at the site.
1.8.4.4.4 All operations involving the fuel are conducted in locations that have been selected to preclude mechanical damage and flooding.
1.8.4.4.5 Loose rods are stored only in RA-series inner metal containers.
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1.8.4.4.6 Fuel is handled in accordance with pertinent provisions of the reactor license and in accordance with written and approved GE procedures.
The procedures shall define the radiation and contamination surveys that are to be performed and the frequency of the surveys.
1.8.4.4.7 Written administrative procedures are jointly prepared and approved by GE and the reactor licensee, to provide for the nuclear and radiation safety of the operations to be performed.
1.8.4.4.8 Records of the operation, including svaluations, procedures used, audits performed, and performance reports are maintained at the Wilmington, North Carolina, plant.
1.8.5 Disposal of Industrial Waste Treatment Products Notwithstanding any requirements for state or local government agency disposal permits, GE is authorized to dispose of industrial waste treatment products without continuing NRC controls provided that eithel of the two following conditions are met:
1.8.5.1 All freestanding liquid shall be removed prior to shipment.
The uranium concentration in the material shipped for disposal shall not exceed 30 pCi per gram after all freestanding liquid has been removed.
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4 The licensee shall possess authorization from appropriate state officials prior to disposing of the waste material delineated in this snendment.
The authorization shall be available for inspection at the Wilmington facility.
1.8.5.2 The uranium concentration in the material shipped for disposal only at the RCRA hazardous waste burial facility in Pinewood, S.
C.
(licensed by the State of South Carolina), shall not exceed 250 pCi per gram of uranium activity, of which no more than 100 pCi per gram shall be soluble.
The min 1 mum burial depth shall be at least four feet below the surface.
1.8.6 Dilution Factor for Airborne Effluents Authorizhtion to utilize a dilution factor to the mearured stack discharges for the purpose of evaluating the airborne radioactivity at the closest site boundary of stack discharges from the uranium processing facilities.
For purposes of control, this dilution factor shall be no greater than 100.
For other purposes, specific dilution factors, which consider dispersion model parameters, may be calculated and used.
1.8.7 Monitor System Exemption Authorizstion for exemption from the requirements of 10 CFR 70.24 for each area in which there is not more than:
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4 A quantity of finished reactor fuel rods equal to or less than 45% of a minimum critical number under conditions in which double batching is credible, or equal to or less than 75% of a minimum critical number under onditions in which double batching is not credible, or The number and type of finished reactor fuel rods and/or assemblies authorized for delivery to a carrier for transport as a Fissile Class I shipment in the model RA-series shipping package described in package certificate USA /4986/AF, without limit on the number of such stored containers, provided the storage locations preclude mechanical damage and flooding, or The quantity of uranium authorized for delivery to a carrier for transport as i Fissile Class I package when fully packaged as for transport according to a valid NRC authorization for such packages without limit on the number of such packages, provided storage locations preclude mechanical damage and flooding, or Arrays of finished reactor fuel rods and/or assemblies in any of the inner metal containers of the RA-series shipping package described in package certificate USA /4986/AF,'under stcraie conditions described in Chapter 1.8.4.3.
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1.8.8 Incinerator Operation Authorization, pursuant to 10 CFR 20.302(a) and 10 CFR 20.305, to treat or dispose of waste and scrap material containing special nuclear material by incineration.
1.8.9 Posting For those areas within the Controlled Access Area in which radioactive materials are processed, used, or stored, where it is deemed impractical to label individual containers pursuant to 10 CFR 20.203(f), a sign stating "Every container in this area may contain radioactive material" shall be posted.
1.8.10 Uranium Recycle Enrichment Control Maximum enrichment in the Uranium Recycle (UPMP) operation shall not exceed the minimum U825 enrichment approved by the nuclear safety function for any Uranium Recycle process.
1.8.11 Sanitary Sludge Accumulation Authorization to accumulate treated sanitary sludge containing trace amounts of uranium, in the sanitary sludge land application area pending final disposal.
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1.8.12 Transfer of Calcium Fluoride (CaF,) Test Quantities Authorization to transfer test quantities of CaF,'to potential buyers for the purpose of their examination and evaluation as described in NF&CM's letter dated 3/21/88.
Test quantities may not contain more than 30 pC1 per gram on a dry weight basis and limited to 1 gram U285 at each offsite location.
Test activities and end use of the material will be limited to those that do not allow chendcal separation of the uranium or entry of the nroduct into the food chain.
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GUIDELINES FOR DECONTAMINATION OF FACILITIES AND EQUIPMENT PRIOR TO RELEASE FOR UNRESTRICTED USE OR TERMINATION OF LICENSES FCR BYPRODUCT, SOURCE, OR SPECIAL NUCLEAR MATERIAL U.S. Nuclear Regulatory Commission Division of Industrial and Medical Nuclear Safety Washington, DC 20555 August 1987 LICENSE SNM-1097 OATE 5/16/88 Page DOCKET 70-1113 REVISION 21 I-1.21
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The instructions in this guide. in conjunction with Table 1. specify t'he radionuclides and radiation exposure rate limits which should be used in decontamination and survey of surfaces or premises and equipment prior to abandonment or release for unrestricted use. The limits in Table 1 do not apply to premises, equipmefit, or scrap containing induced radioactivity for which the radiological considerations pertinent to their use may be different.
The release of such facilities or items from regulatory control is considered on a case-by-case basis.
1.
The licensee shall make a reasonable effort to eliminate residual contamination.
2.
Radioactivity on equipment or surfaces shall not be covered by paint.
plating, or other covering material unless contamination levels, as determined by a survey and documented, are below the limits specified in Table 1 prior to the application of the covering. A reasonable effort must be made to minimize the contamination prior to use of any covering.
3.
The radioactivity on the interior surfaces of pipes, drain lines, or ductwork shall be determined by making measurements at all traps, and other a#ropriate access pints, provided that contamination at these locations is likely to be repicsentative of contamination on the interior of the pipes, drain lines, or ductwork. Surfaces of promises, equipment, or scrap which are 1thely to be contaminated but are of such size, construction. or location as to make the surface inaccessible for purposes of measurement shall be presumed to be contaminated in excess of the limits.
4 Upon request, the Commission may authorire a licensee to relinquish possession or control of premises, equismient, or scrap having surfaces contaminated with materials in excess of the limits specified. This may include, but would rot be limited to, special circumstances such as razing of buildings, transfer of promises to another orpnization continuing work with radioactive materials or conversion of facilities to a long-tern storage or standby status. Such requests rust:
a.
Provide detailed, specific information describing the presises, equipment or scrap, radioactive contaminants, and the nature, estent, and degree of residual surtsca contamination, b.
Provide a detailed health and safety analysis which reflects that the residual amounts of materials on surface areas, together with other considerations such as prospective use of the premises, equipment, or scrap, are unlikely t' result in an unreasonable risk to the hesith and safety of the publir LICENSE SNM-1097 DATE 5/16/88 Page 00CKET 70-1113 REVISION 21 I-1,22 f
5.
Prior to release of premises for unrestricted use the Itcense shall s.ake a comprehensive radiation survey which establishes ti.at contamination is within the limits specified in Table 1.
A copy of th". survey report shall be filed with the division of Industrial and Noical Nuclear Safety.
U. 5. Nuclear Regulatory Commiss'en Washi ston, DC 20555, and also the Administrator of the NRC Regional Office having jurisdiction. The report should be filed at least 30 days prior to the planned date of abandonment.
The survey report shall:
a.
Identify the premises.
b.
Show that reasonable effort has been made to eliminate residual contamination, c.
Describe the scope of the survey and general procedures followed, d.
State the findirgs of the survey in units specified in the instruction.
Following review of he report, the NRC will consider visiting the facilities to confim the survey.
LICENSE SNM-1097 DATE 5/16/88 Page DOCKET 70-1113 REVISION 21 i
I-1.23
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TAStl 1 ACCIpTA8tt $URFACI CONTAMitt* TIM LEYtt$
agnoyagg b e f Is0CtlDE$8 AVERAC(b c f MA11 W d f U. net. U 239. U 238, and t
15.000 dpa e/100 c,2 1.000 des e/100 cm2 associated decay products 5.000 dpa e/100 cu TraAsurantes. Ra.226. ta.228.
100 dom /100 c,2 300 dpa/100 ce!
20 dpa/100 co2 Th 230. Th.228, ps.231 Ac.227.1 125.1 129 Tb net. T b 232 $r.90 1000 don /100 c,2 3000 dpm/100 cm2 200 dpe/100 c,2 Ra.223. Re.224. U.232.1 126, 1 131. 1 133 Seta garra emitters (nuclides 2
1000 dpa ar/100 cat with decay rodes other than 5000 dpa sy/100 cm2 15.000 den sy/100 cm alpha esission or spontaneous fission) except $r 90 and others noted above.
k fttlag 4 here surf ace conteetnation by bcth alpha. and bets.gaans.eettting nuclides entsts. the limits established for alpha. and Wnucif des shch.1d apply independently.
bas used in this table, dem (dtsintegratices per minute) means the rate of esistio For objects of less surface area, the average cMealvrements of everage contaminant should not be averaged over more than 1 square arter.
should be serived for each such objett.
fhe sanimum contamination level applies to en area of not more than 100 col.
d
'The amvat of removable radioactive material per l'00 ca! of surface area should When removante centsetnation on objects of less surface area is detersined, the pertinent levels should be r known efficiency.
proportionally and the entf re surface should be utped.
The average and manian radtation levels associated with surface contamination resulting from beta.gamme te 1 cm and 1.0 erad/hr at 1 cm. respectively, measured through not more than 7 mil 11 grams per square centim f
0.2 mead /hr at total absorber.
I LICENSE SNM-1097 DATE 5/16/88 Page 00CKET 70-1113 REVISION 21 I-1,24 i
i
i reviewed by the committee.
Such reports shall be retained for at least two years.
The committee shall hold at least four meetings each calendar year with a maximum interval of 180 days between any two consecutive meetings.
2.3.2 Operational Radiation Safety Committee The objective of the operational Radiation Safety Committee is to improve fuel manufacturing operations so as to affect visible improvements in employee radiological exposures and potential health / safety hazards.
The Committee meets monthly to maintain a continual awareness of the status of containment projects, performance measurement and trends, and the current shop operations radiation safety conditions.
The maximum intervals between meetings shall not exceed 60 days.
A written report of each operational Radiation Safety Committee meeting shall be forwarded to cognizant Area l
Managers and the manager of the regulatory compliance i
function.
The report shall be forwarded within 15 working days of the meeting.
Records of the committee proceedings shall be maintained for two years.
The Committee shall consist of engineering, shop operations, maintenance, radiation safety function, radiation protection function and quality control management or senior contributors.
LICENSE SNN-1097 OATE 5/16/88 Page DOCKET 70-1113 REVISION 21 I-2.8 em
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safety functions.
They shall assure that these procedures are made readily available to foremen, operators and other concerned personnel through posting of limits, training programs and other appropriate VTitten notifications.
l The radiation protection program is designed to establish and maintain a comprehensive set of written instructions for radiation health and safety practices so as to maintain occupational radiation exposures at levels as low as reasonable achievable.
Such I
instructions are reviewed by the radiation safety function prior to issuance by appropriate managers.
l Subsequent procedure review frequencies are.shown below:
Reviewing &
Approving Review Functional Document Frequency Manager Operational (Irocess When Radiation Safety l
Requirement changed *
& Criticality
& Operator Documents)
Safety
. Radiation Protection Annually 8 Radiation Safety (Nuclear Safety
& Criticality Instructions)
Safety Environmental Every 2 Environmental (Environmental Years 8 Protection Protection l
Instructions) 2The radiation safety and c'riticality safety portions of these operator procedures are reviewed and updated by the radiation safety and criticality safety functions when varranted based on process related i
facility change requests.
i 8 Annually means a maximum interval of 13 months.
)
l 8Every 2 years means a snaximum interval of 26 months.
l l
l l
LICENSE SNN-1097 DATE 5/16/88 Page DOCKET 70-1113 REVISION 21 I-2.15 l
l 1
F The activity release data are also accumulated and reported on a semi-annual basis to the Nuclear Regulatory Commission.
5.1.1.4 Lower Limits of Detection, Calibration and Standardization of Measurements An alpha sensitive detector is used to determine the activity on the stack sample filters.
The system is calibrated for gross alpha using a standard traceable to the National Bureau of Standards.
This system provides a lower limit of detection of activity at the site boundary within the requirements of Appendix B of 10 CFR 20.
Background and efficiency checks are performed each operating shift.
The minimu.' detection limit for airborne effluent concentrations is 1 x 10-28 pC1/cc at the stack which l
equates to approximately 1 x 10-2* pC1/cc at the site boundary using a dilution factor of 100.
This lower limit of detection is lower than the limit identified in Regulatory Guide 4.16, March 19"8, and is 1 10% of the value 3 x 10-28 pCi/cc, from Table II of Appendix B l
to 10 CFR 20.
4 I
5.1.2 Liquid Effluents l
The liquid waste streams containing uranium from the fuel manufacturing operations are segregated as nitrate l
waste, fluoride waste and rad waste.
This separation LICENSE SNM-1097 OATE 5/16/88 Page DOCKET 70-1113 REVISION 21 I-5.6 i
6.4.4 Commitment of Compliance Full compliance with the commitments in Section 6.4 will be achieved by 1987 year end.
6.5 Dry Conversion Process UF, gas flows into a steam hydrolysis reactor and is contacted with dry superheated steam.
An instantaneous and exothermic reaction assures essentially immediate and complete conversion.
The following table describes the safety interlocks and frequency of safety assurance checks.
Dry Conversion (Steam Hydrolysis) Control Table Frequency Safety Interlocks Calibration 1 Function Tests a Steam
- Negative pressure Semi-Annually Annually Hydrolysis prevents UF, and Reactor steam backflow a
Pressure
- Automatic UF, shutoff if vacuum is lost
- Main vacuum control valves fail open
- Feed shutoff valves fail closed 1If the process hcs been shut down and more than six months have elapsed since the previous calibration, the principal safety interlocks shall be calibrated prior to startup.
8If the process has been shut down and more than one year has elapsed since the previous functional test, each principal safety interlock shall be functionally tested prior to startup.
LICENSE SNM-1097 DATE 5/16/88 Page DOCKET 70-1113 REVISION 21 i
I-6.3
r Dry Conversion (Steam Hydrolysis) Control Table - Continued Frequency 7
Safety Interlocks Calibration 1 Function Tests 8 l
- Negative pressure assures containment l
of HF and uranium f
Steam to
- Stoichiometric Semi-Annually Annually UF, Flow controls close the Ratio UF, valve Superheated
- Low superheated Semi-Annually Annually Steam steam temperature Temperature shuts off UF, supply t
- Lower superheated t
steam temperature shuts off steam supply Steam
- Two reactor Semi-Annually Annually i
Hydrolysis temperature l
Reactor detectors Temperature
- Low reactor j
temperature shuts 1
off UF, supply
- Lower reactor
=
temperature shuts off steam supply j
i SIf the process has been shut down and more than six months have a
elapsed since the previous calibration, the principal safety interlocks shall be calibrated prior to startup.
8If the process has been shut down and more than one year has elapsed since the previous functional test, each principal safety interlock shall be functionally tested prior to startup.
LICENSE StN-1097 DATE 5/16/88 Page DOCKET 70-1113 REVISION 21 I-6.4
s Dry Conversion (Steam Hydrolysis) Control Table - Continued Frequency Safety Interlocks
_ Calibration 1 Function Tests 8 Primary
- Two temperature Semi-Annually Annually Filter detectors Temperature
- Low filter temperature shuts off the UF, supply o Lower filter a
temperature shuts off steam supply Low or high dif-Semi-Annually Annually Secondary Filter ferential pressure Differential will shutdown the Pressure reactor, primary filter and the dry air purges High level cuts off Semi-Annually Annually HF Stripping UF, feed, steam Column supply and DI water Level 8 with valves in closed position LIf the process has been shut down and more than six months have elapsed since the previous calibration, the principal safety interlocks shall be calibrated prior to startup.
8If the process has been shut down and nore than one year has elapsed since the previous functional test, each principal safety interlock shall be functionally tested prior to startup.
8 Located in an isolated room constructed of HF resistant material.
Entry is not permitted when system is operating.
LICE','id SNM-1097 DATE 5/16/88 Page D0t.dT 70-1113 REVISION 21 I-6.5
A V
Dry Conversion (Steam Hydrolysis) Control Table - Continued Frequency i
Safety Interlocks Calibration 2 Function Tests 8
'f MF e Low flow trip Semi-Annually Annually Stripping point shuts off Column DI UF, feed Water
- DI water flow Flow
- control valve fails in open position Defluor-
- Pressure below a Semi-Annually Annually t
inator Feed low trip point Preparation shuts off defluor-l Unit inator hydrogen l
l with valve in closed position
- If the process has been shut down'and more than six months have elapsed since the previous-calibration, the principal safety interlocks shall be calibrated prior to startup.
8If the process has been shut down and more than one year has elapsed since the previous functional test, each principal safety interlock shall be functionally tested prior to startup.
- Located in an isolated room constructed of HF resistant material.
Entry is not permitted when system is operating.
L I
f LICENSE SM4-1097 DATE 5/16/88 Page DOCKET 70-1113 REVISION 21 I-6,6 j
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