ML20010B074
| ML20010B074 | |
| Person / Time | |
|---|---|
| Site: | 07000824 |
| Issue date: | 07/23/1981 |
| From: | Olsen A BABCOCK & WILCOX CO. |
| To: | Crow W NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| 19421, NUDOCS 8108130292 | |
| Download: ML20010B074 (53) | |
Text
i f$~ h opg Babcock & Wilcox
"**E*u"rg'n.*.'.*.'re'n*c*."n'tSE "
yn a McDermott company P.O. Box 1260 O
Lynchburg, Virginia 24505 (804) 384-5111 g
- tocm, l
6 fuly 23,1981 ) *Y, II P:
AUG O J
6 jgg,
- to,y l
ir. W. T. Crow, Section Leader N
Uranium Process Licensing Section y
g u
l Uranium Fuel Licensing Branch p
Division of Fuel Cycle and Material Safety a"
U.S. Nucleizr Regulatory Commission y
hashington, DC 20555 m..
Reference:
(a) BAW-381, Demonstration and Conditions for LiceWse SNM-778, December 1978.
(b) License SNM-778, Docket 70-824
Dear Mr. Crow:
The Lynchburg Research Center reorganized some of its technical sections resulting in changes in the titles.
Figure A-4, page A-47 of Reference (a) was affected. To make this change it is necessary to amend Reference (b). A check for $150M is attached to cover the fee for an administrative amendment specified in 10 CFR 170.31, Category 1.F.
I have taken this opportunity to make a few needed corrections in the demonstration sections in Reference (a). These changes are attached along with a descripthn of each change and an instruction sheet for entering the replacement pages in Reference (a).
Please contact me if you have any questions in this regrd..d m
Very truly yours, g BABC0CK & WILC0X rb h
u4 9
A. F. Olsen h
Senior License Adainistintor un ATTACHMENTS
-g I Wh
/
Applicant... f............
Check 'io. 0.
ry.......g.y' f
Q amouny Fea Cate Jf.
s fype of i'oc.
--}
O
.T.
Date Check Rec'd.,
.M,... '-
N 5
Received By.
l h
1 3.I/21
.ru S-f 8108130292 810723 ~
YCo'/ '/fl%
/
{DRADOCK07000g
6 6
Description of Changes Description Page 2-8 Figure 2-1 Revised 2-9 Figure 2-2 Revised 3-2 Changed "LRC-151 " Work Order" to "LRC-229 " Facilities Work Order Form".
3-3 Changed "capabile" to " capable".
3-7 In'5 3.4.3.6, line 5, removed "of" between " block" and " Wall".
In 5 3.4.3.8. line 3, changed "second" to "first".
3-12 5"3.5" Building C removed.
3-13 Added 5"3.5 Building C" In 5 3.5.2, changed " Analytical Chemistry" to "Radioanalytical Chemistry".
In 5 3.5.2, line 2, changed " analytical chemistry" to "Radioanalytical rhemistry".
3-14 The title of 53.5.3, " Development" changed to " Technology".
In 5 3.5.3 line 1. changed " development" to " technology".
In the title of 5 3.5.4, changed " Fuel Materials" to "Special Projects".
l In 5 3.5.4, line 1. changed " fuel acterials" to "Special Projects".
3-33 Form 151 replaced with form 229.
4-10 Changed "arrary" to " array".
4-11 In 5 4.4.4.2, line 2, changed "that" to "to".
4-14 In 5 4.4.4.5.2. A, last line, changed "4.4.5.2.B" to "4.4.4.5.2.B".
4-18 In paragraph 1, line 7, changed "vaid" to " valid".
4-20 In 5 4.4.7, added reference "(A.9.5.10)".
4-21 In 5 4.5.1, Item 1, changed "surveilance" to " surveillance".
4-22 In 5 4.5.2.1. line 6, "subjet" changed to " subject".
5 4.5.5.2 corrected to 5 4.5.2.2.
5 4.5.5.2.1 corrected to 5 4.5.2.2.1.
4-23 First paragraph, line 5, "randon" changed to_ " radon".
f I
l
Description of Changes Page Description 5 4.5.2.2.2.1. line 8, "randon" changed to " radon".
4-26 Second paragraph, line 2, "Reiview" changed to " Review".
4-28 First paragraph, line 7, " mater" changed to " material".
4-29 In the second line under " ionization Chambers" " sever 1" changed to "several".
4-32 5 4.5.3.2, line 1 "25" changed to "10".
In line 5 of the note, "bisit" changed to " visit".
4-38 First paragraph, line 7, "phycis" changed to " physics".
4-39 In the second line under " Annual Retraining" "attent" changed to " attend".
4-40 5 4.5.10, item 1, line 3, "or" changed to "of".
1 4-41 5 4.5.12, item 1.a, "LRC-151 " Work Order forms", changed to "LRC-229
" Facilities Work Order Forms".
4-42 5 4.5.13, paragraph 2, line 4, " imminently" changed to " eminently".
4-43 Paragraph 6, line 1, " Respirator" changed to " Respiratory".
4-45 Paragraph 2, line 2, "consier" changed to " consider".
A-47 Figure A-4 Revised
____-.-_-._m_
a-INSTRUCTION SHEET BAW-381 Demonstration and Conditions for License SNM-778, Revision 14 and Amendment 1, July,1981.
Replacement Page Take Out Page Page Rev Date h
Rev Date 2-7 1
9/79 2-7 1
9/79 2-8 14 7/81 2-8 2
10/79 2-9 14 7/81 2-9 2
10/79 3-1 0
12/78 3-1 0
12/78 3-2 14 7/81 3-2 0
12/78 3-3 14 7/81 3-3 0
12/78 3-4 0
12/78 3-4 0
12/78 3-7 14 7/81 3-7 1
9/79 3-8 1
9/79 3-8 1
9/79 3 1 9/79 3-11 1
9/79 3-12 14 7/81 3-12 1
9/79 3-13 14 7/81 3-13 1
9/79 3-14 14 7/81 3-14 1
9/79 3-33 14 7/81 3-33 1
9/79 3-34 1
9/79 3-34 1
9/79 4-9 1
9/79 4-9 1
9/79 4-10 14 7/81 4-10 1
9/79 4-11 14 7/81 4-11 2
10/79 4-12 2
10/79 4-12 2
10/79 4-13 2
10/79 4-13 2
10/79 4-14 14 7/81 4-14 2
10/79 4-17 5
1/80 4-17 5
1/80 4-18 14 7/81 4-18 3
11/79 4-19 2
10/79 4-19 2
10/79 4-20 14 7/81 4-20 2
10/79 4-21 14 7/81 f
4-21 1
9/79 4-22 14 7/81 4-22 1
9/79 4-23 14 7/81 4-23 1
9/79 4-24 3
11/79 l
4-24 3
11/79 4-25 1
9/79 4-25 1
9/79 4-26 14 7/81 4-26 1
9/79 4-27 1
9/79 4-27 1
9/79 4-28 14 7/81 4-28 1
9/79 4-29 14 7/81 4-29 1
9/79 4-30 1
9/79 4-30 1
9/79
INSTRUCTION SHEET -- BAW-381 Demonstration and Conditions for License SNM-778, Revision 14 and Amendment 1, July, 1981.
Replacement Page Take Out Page Page Rev Date Page Rev Date 4-31 1
9/79 4-31 1
9/79 4-32 1
9/79 4-32 14 7/81 4-37 1
9/79 4-37 1
9/79 4-38 14 7/81 4-38 1
9/79 4-39 1
9/79 4-39 14 7/81 4-40 14 7/81 4-40 1
9/79 4-41 14 7/81 4-41 2
10/79 4-42 14 7/81 4-42 1
9/79 4-43 14 7/81 4-43 3
11/79 4-44 1
9/79 4-44 1
9/79 l
4-45 14 7/81 4-45 1
9/79 4-46 1
9/79 4-46 1
9/79 A-47 2
10/79 A-47 14 7/81
?
l l
[')
2.2.9 License Administrator V
The license administrator shall have at least a BS degree in science or engineering and three year's experience in nuclear technology or an AS degree in science or nuclear technology and 12 year's experience in nuclear technology. He is responsible for administering the LRC licenses. The administrator is the primary liaison with the NRC and other federal, state, and local agencies regarding nuclear mattars. He is the coordinator of the Safety Review Committee and Audit Subcommittee and represents management on both. He is also coordinator of the facility supervisors. The license administrator reports to the Manager of Accounting and Administrative Services.
2.3 FACILITIES DEPARTMENT Facilities include health physics, industrial safety, design and drafting services, a fully equipped machine shop, instrument and equipment maintenance, modification and installation services. The machine shop and drafting personnel occupy more than 4700 square feet of floor space in Building B.
Maintenance, modification, and installation services are also p~,
()
provided by the Facilities Department. An instrument service group main-tains a fully equipped electronics shop to install, maintain, modify, and calibrate the electronic, health physics, and analytical equipment neces-sary for safety and for a satisfactory SNM safeguards program. This group maintains all of the electronic equipment used for routine operation of the LPR, the critical assembly, the precision instrumental and analytical equip-ment, and special R&D apparatus. A separate group provides mechanical, plumbing, electrical maintenance, and installation services. Included are the erection and installation of experimental acckup facilities and assistance
{
in installing pilot plants and special research and development facilities.
l License No.
SNM-778 Docket No.70-824 Date September, 1979 Page 2-7 Amendment No.
Revision No.
1 Babcock & Wilcox
s r.
a n
e e
DIRECTOR s
3 cL T. C. Engelder 3
e Z
a
?
I I
I I
Z FACILITIES PURCHASING PERSONNEL ADMI4 i
RVICES Is C. E. Bell K. A. Gondert J. P. Doran J. R. Parsall N
b Manager Manager Manager Manager o"n k.
I l
7 35 O
MATERIALS & CHEMISTRY LABORATORY SENIOR SCIENTIST SYSTEMS DEVELOPMENT LABORATORY n
E. D. Lynch, Manager C. S. Caldwell A. E. Wehrmeister, Managte g
y n
~*
l o
CERAMICS NOUDLSTRUCTIVE METHODS
[
s
& DIAGNOSTICS D
G' ' "9
"'ction Manager Z
Se A. E. Holt Z
?
Section Manager k
O 8
w CHEMICAL & NUCLEAR
~
r
?
ENGINEERING PROCESS CONTROL y
e R. H. Lewis R M NTATION P
Section Manager R. L. Currie y
Section Manager EN O
IRRADIATED MATERIALS 7
0 TECHNOLOGY SYSTEMS DESIGN ch) e H. H. Davis OI e
'O Section Manager J. M. Kerr e
4 X
5 Section Manager 90 hUCLEAR MATERIALS TECHNOLOGY w
G. S. Clevin;cr k
Section Manager
_=
~
t9X O
G G
Y I
hb q
t L'
t 1 z ';
t "..
i t :
. c,..i.
- 2
-.a.
a m
W E2 C u G 1:3.i a
a.
a, u
v Y ': : 2
-.ti, i is 22
'9
?
- .3
. o 51.
o w
u 4
2
,, i
- x
,n. ".
~
4-2 g
,s, o
-=
4, a g
jJ j a :i C
Ou 30 4
-n
=:
5, o,
r e,
s 5
u, 3,
e 3
,ga y
=
e 2 2 w =
=
d)
[
e
- 2
-i 3
J 223 4 t
u l
7
~
5 2
e8s w
9
,a
=u u, s, 1:
.x A
i 4
3 4 ::
~.
xa w
e, v a
Q t.
4..
la b
E
~,.,
a.
..,.f
'2 i na
.4 2:
License N o.
SNM-778 Docket N o.70-824 D ate J'21"' 1981
-)
Amendment N o.
Revision N o.
14 Page 2-9 Babcock & Wilcox
,-,y
.,-.-,..-,--.-_,,-,---..,.,.e,--..
,,.a
/O C/
3.
DESCRIPTION OF SITE, FACILITIES, AND LICENSE ACTIVITIES 3.1 SITE The description of the Lynchburg Research Center (LRC) site is presented in " Environmental Report, Lynchburg Research Center", dated December 1978. Figure 3-1 shows the site layout.
3.2 FACILITIES AND LICENSED ACTIVITIES, GENERAL The LRC 1s a highly integrated facility built to develop, test, and examine nuclear reactor cr res and to develop overall fuel cycles.
It is comprised of Buildings A, B, and C (Fig. 3-2).
Building D, a recent addition to the LRC, is used for nonnuclear research and development, and also houses the Center's administration.
Building A has the staff and facilities to conduct full-scale critical experiments on power reactor cores, exponential experiments, and small lattice experiments. The nuclear characteristics of production cores (s,)
and control rods can be determined before shipment to the reactor site.
Building A has been active in the development of incore instrumentation systems and nuclear and nonnuclear computerized proce=s control systems.
Building B provides facilities for post-irradiation examination of commercial fuel assemblies, fuel rods, and reactor components. This building houses metallurgy laboratories, a radiochemistry laboratory, coanting laboratory, health physics facilities, a scanning electron micro-scopy laboratory and a fully equipped machine shop.
Building C provides facilities for development of fuel processing methods through the pilot plant stage, investigating improvements in present fuel cycle tecnnology, handling high alpha emitters, and x-ray diffraction and emission studies.
3-1
-~s License No.
SNM-778, Docket No.70-824 Date December 1978 Page
\\- /
Amendment No.
Revision No.
Babcock & Wilcox
a 3.3 SUILDIUG A 3.3.1 General Building A provides 20,000 square feet of floor area. One Critical Experiment Facility (CX-10) and the Lynchburg Pool Reactor (R-47) are located in the building. Both are licensed pursuant 10 CFR 50 and license SNM-778 does not provide for the possession and use of licensed material in either reactor. However, it does allow for the possession and use of licensed material in both areas outside the reactor.
Building A is utilized for computer studies and development, instru-ment development and testing of ceramic linings in pressure vessels, in d
addition to reactor studies. Very little licensed material is utilize outside the reactor areas.
3.3.2 1
.. sed Material Handling i
Licensed material has been handled throughout Building A in its more than 20 years..? operation. This work has been performed on a project basis. At the conclusion of a project, each area has been cleaned for I
release for unrestricted use before the start of a new project. Bay 1 l
l (Fig. 3-3) is a good example of this process.
This area was constructed in 1956 to house the"first privately-l owned critical experiment, the CX-1.
Later the CX-19 facility shared this i
bay.
In 1971, it became evident that these facilities were no longer useful.
[
They were completely decontaminated and decommissioned in 1972. Today this bay is an unrestricted area and houses a project that utilizes no licensed material.
New projects invc1ving the use of licensed material are reviewed by the facility supervisor and the Safet.y Review Committee.
3.3.3 Facility Change Changes and modifications to buildings, exhaust ventilation systems, gas supply systems, emergency electrical systems, etc., are requested on Form LRC-229 " Facilities Work Order Form" (Fig. 3-6). All work orders are forwarded to License No.
SNM-77{
acket No.70-824 Date July, 1981 Page 3-2 Amendment No.
Revision No. 14 Babcock & Wilcox
[/
i
(_
the maintenance supervisor. The maintenance supervisor determines if the request involves a facility change. If a facility change is involved, the work order is forwarded to the facility supervisor.
It is the facility supervisor's responsibility to determine that all safety and licensing considerations have been addressed and if the request must be approved by the Safety Review Committee. Space is provided on the form for the approval signatures of health physics, the industrial safety officer, and the facility supervisor.
Completed forms are kept on file by the maintenance supervisor and are audited once a month by the health rhysics group.
3.3.4 Ventilation Figure 3-7 shows the present ventilation in Building A, except for the building conditioned air system. Licensed material that is handled in t.e building is sealed sources, fuel rods and elements, irradiated hard-wr.e and flux measuring foils and wires. Material that presents a potential release hazard is not presently handled. C snges to the ventilation system,
(}
other than the conditioned air system, are 1, iewed b) the facility super-visor and the Safety Review Committee (Sht, SRC approval is also required on new projects utilizing licensed material 3.3.5 Criticality Accident Alarm System The criticality accident alarm system is a Tracer-Lab RMI-110. Four detect 2rs are capable of activating the building evacuation alarm (ref.
I Fig. 's-3). Detector C-1 is located at the ceiling of the first floor; C-2 is located in the Penthouse above the first floor; C-3 is located at the Bay 2 absolute filter; and C-4 is located on the LPR console. The location, setpoint and function of the system meets the requirements of 10 CFR 70.24 (a)(2), except detector C-3.
This detector is installed to detect an inadvertent criticality accident in Bay 2.
Therefore, it is set to alarm at a radiation level twice that expected from the critical experiment or 750 mR/hr.
f-~s License No.
SNM-778 Docket No.70-824 Date July. '981 Page 3-3 o
\\
\\~ /
Amendment No.
Revision No.14 Babcock & Wilcox
The interior walls of Building A are constructed of cinder block, g
which offers little shielding against radiation. The two bays, physics laboratory and neutron noise analysis laboratory (Fig. 3-3) are the only areas that offer appreciable shielding. The criticality monitors are so located that an accident, as described in 10 CFR 70.24 (a)(2), occurring in any material handling area would be detected by at least one of these monitors. As material handling areas may be redefined from time to time, the indicated locations should tot be considered permanent. Monitors may be added, removed or relocated eith the approval of the Sr ty Review Committee.
The system is normally powered by building powec. Emergency power is supplied by a battery-powered invertastat.
l l
.4 BUILDING B l
3.4.1 General l
l i
Building B, which occupies 35,000 square feet of floor space, includes a four cell Hot Cell facility with its a.
ted operations area, Hot Instrument Repair Shop, Crane and Cask Handlins a, and a transfer canal. L cated also in Building B is a Metallurgy Labo,ratory, capsule assemb'y area, and an experimental pool located within a metal containment.
3.4.2 Physical Metallurgy Laboratory The Metallurgy Laboratory has equipment for structural e.221 nations on a macroscopic 6nd cicroscopic scale. Facilities are available for all i
i metallography preparations and examinations utilizing light-microscopy. A hot stage metallograph is available for mic~oscopic examination of Laterials t
at high temperatures and in controlled atmospheres. The laboratory is equipped for determining the mechanical and physical properties of metals and for evaluating the affects of heat treatment on the properties of metals. An industrial x-ray unit is also available to this laboratory.
License No.
SNM-778 Docket No.70-824 Date December 1978 Page 3-4 Amendment No.
Revision No.
Babcock & Wilcox
(m) radioactive materials, poolside manipulations, and the raising and lowering
'O of shielded shipping containers.
3.4.3.6 Demineralizer Equipment for maintaining water quality for the transfer canal and storage pool and the containment pool is located below grade in the area designated as " Storage" in the upper left hand portion of Figure 3-4.
The area containing this equipment is separated from the storage area by a concrete block wall and a locked door. This room is considered a high radiation area and access is controlled in com-pliance with 10 CFR 20.203(c)(2).
~ e water treatment equipment is compriced of filters and ion exchange -vlumns.
Water from the two pools is circulated through sep-arate systems. Ion exchange resins are disposed of as dry waste when depleted. Drains from the water treatment equipment empty into a sump which is pumped to the liquid waste disposal facility.
3.4.3.7 Fracture Mechanics Area (Fig. 3-4) - The fracture mechanics area is located on the second floor of Building B, adjacent to the hot cell r"'N operations area. This area contains a closed-loop electrohydraulic load t\\- 'I frame, an impact tester and a fatigue precracker.
3.4.3.8 Scanning Electron Microscope (Fig. 3-4) - The scanning electron microscopy (SEM) laboratory is located adjacent to the hot cell operations area on the first floor of Building B.
This laboratory consists l
of a sample preparation room and the SEM room. This research tool is used for both radioactive and nonradioactive materials.
3.4.4 Capsule Assembly Room (Fig. 3-4)
The capsule assembly room is used to prepare reactor vessel surveil-lance capsules, in-reactor experiments, and for preparation of tubing samples for pressure testing.
License No.
SNM-778 Docket No.70-824 Date July, 1981 Page 3-7 Amendment No.
Revision No.
14 Babcock & Wilcox
3.4.5 Radiochemistry Laboratory (Fig. 3-4)
The radiochemistry laboratory contains several chemical fume hoods, a perchloric acid hood and other equipment normally found in a well equipped laboratory. This laboratory is located in the controlled-access portion of Building B.
This facility is used for handling radioactive components including small quantities of irradiated fuel.
Samples are dissolved and prepared for counting to determine the radicactivity or for isolation of specific elements for further analyses.
3.4.6 Counting Laboratory (Fig. 3-4)
The counting laboratory contains several high resolution gamma spectroscopy systems coupled to computers for data processing. A liquid scintillation system is used for spectroscopy of low energy beta emitters.
Cross counting and spectroscopy are performed on alpha and beta emitting elements.
3.4.7 betallurgy Laboratory - (Fig. 3-4)
The metallurgy laboratory has equipment for structural examinations on a mtcroscopic and microscopic scale. Facilitie.s are available for all metallography preparations and examinations utilizing light-microscopy. A hat stage metallograph is available for microscopic examination of materials at high temperatures and in controlled atmospheres. The laboratory is equipped for determining the mechanical and physical properties of metals and for evaluating the effects of heat treatment on the properties of metals. An industrial x-ray unit is also available to this laboratory.
3.4.8 Ventilation and Materials Handling (Fig. 3-12)
Building B is air-conditioned, and the overall ventilation system has been designed so that the normal flow of air is toward areas of potential contamination in order of severity, the lowest pressure being within the hot cell.
License No.
SNM-778 Docket No.70-824 Date September,1979 Page 3-8 Amendment No.
Revision No.
1 O
Babcock & Wilcox
The hot cell is maintained at the lowest relative pressure, thus' ensuring that any contamination is contained. The design criteria specify b) a pressure differential of 0.6 inch water across the cell face (between
(_
the operations area and the cell interior) and a linear velocity of 100 cfm when either the door or the roof slab is open. Administrative procedures prevent the possibility of both major openings being open at the same time.
The volume of air is sufficient to remove the heat generated by lighting plus 7.5 kW from other sources and to maintain ambient temperatures below 120 F during normal operating conditions. Fire-resistant Fiberglas pre-filters are located inside the cell.
Air from the hot instrument repair shop is exhausted through a single fan and absolute filter and is discharged to the suction side of the main stack blower. Air from the cask handling area is recirculated through a single absolute filter to the building supply system. Air exhaust from the isolation area and the hot cells passes e.hrough two series banks of three fire-resistant absolute filters enclosed in a meter filter housing.
All ducting between the hot cells and the filters is of steel construction.
A fire damper operated by a fusible link is located upstream of the filter bank.
i
\\- '
Air flow is produced by dual exhaust fans, connected in "and/or" circuitry, each rated at 1600 cfm at a 5.8-inch head.
Filters and fans are located on the roof of the isolation area. The design criteria require that either fan must be able to maintain the 0.6-inch of water pressure differ-ential across the cell walf. when all accesa cpenings are closed. When the pressure differential drops below this value, the second fan is automatically energized and indications are given on a panel in the operations area. A low-pressure differential alarm and manual bypass of the automatic system permit proper operacian in the event that the automatic system fails. A pressure differential alarm and indicator across the absolute filter bank are actuated by blockage of the filters and are also tied to the previously mentioned alarm panel in the operations area. One station of the area monitoring system is located near the filter bank and gives remote indicatien of any activity buildup in the filters.
License No.
SNM-778 Docket No.70-824 Date September, 1979 Page 3-11 Amendment No.
Revision No.
1 Babcock & Wilcox
The exhaust from the cell ventilation system is isolated from other portions of the building system by backdraft dampers until it enters the inlet to the main stack blower. At a point approximately 25 feet up the stack, a sampling station constantly monitors the exhausted air.
Two separate systems supply emergency power for ventilation: one system feeds one of the cell exhaust fans, and the other drives the main blower at the base of the stack. The switchover systems for these emergency power supplies are automatic, but there are manual controls in case the automatic systems fail. Emergency power for the main building exhaust fan is controlled from the hot cell operations cree with indicators at that position.
3.4.9 Criticality Accident Alarm System Figure 3-4 shows the location of the system monitors. The system detects gamma radiation and meets the 10 CFR 70.24 (a)(1) requirements.
These monitors are connected to a logic circuit that activates the building evacuation alarm when any two monitors trip. The monitors are connected to the building's energency power system, ensuring alarm activation in the event of a failure of normal site power. Monitors may be removed, added and relocated to meet the Center's experiment programs; therefore, the locations above should not be considered permanent.
Changes in monitor location and their removal shall comply with 10 CFR 70.24 (a)(1) and be approved by the Safety Review Committee.
3.4.10 Facility Change Reference Section 3.3.3.
\\
l 1
l License No.
SNM-778 Docket No.70-824 Date July, 1981 Page 3-12 1
i Amendment No.
Revision No.
14 O
1 Pabcock & Wilcox
l 3.5 BUII. DING C 3.5.1 General Building C provides 20,000 square feet of laboratory, office, and support space. The building was designed for handling multikilogram quantities of plutonium. The present license limit severely restricts the amount of plutonium that can be handled but the facilities for handling large quantities remain. The research and development performed in Building C primarily involves the use of unirradiated source and special nuclear material. Some work involving the use of byproduct is carried out in the facility but it is very limited.
I 3.5.2 Radioanalytical Chemistry The Radioanalytical Chemistry group provides routine analytical chemistry services to other groups at the LRC as well as other divisions of the Company as needed. This effort includes development of new analytical methods, if needed. This group utilizes laboratories 19, 20 and 27 (Fig. 3-5) generally. The following major items of equipment are available in these areas for their use.
1.
X-ray diffraction, 2.
Emission spectrometer.
g
(.s' 3.
Atomic absorption.
4.
Polarograph.
5.
Gas chromatograph.
6.
Spectrophotometers.
7.
Carbon, sulfur, and halides analyzers.
8.
Moisture analyzer.
9.
Differential thermal analyzer.
10.
Subsieve sizer.
Additionally, there are numerous items of equipment needed to do traditional wet chemical analyses and a computer terminal is available for data reduction.
l l
Some of the above equipment is located within glove boxes if it has been or is planned to be used in handling plutonium.
l License No.
SNM-778 Docket No.70-824 Date July, 1981 Page 3-13 Amendment No.
Revision No.
14 Babcock & Wilcox
3.5.3 Process Technology Group.
The process technology group provides process engineering develop-ment and design for the Company in the following nuclear fuel cycle areas.
1.
Fuel conversion.
2.
Fuel, control and moderator materials fabrication.
3.
Scrap recovery.
4.
Effluent treatment.
5.
Waste treatment.
6.
Non-aqueous, non-gaseous coolants.
This group utilizes laboratories 43, 44, and 50 (Fig. 3-5).
The following major items of pilot-plant-scale equipment are available in these areas for their use.
1.
Effluent treatment pilot plant.
2.
Fluid bed.
3.
Reverse osmosis.
Additionally, there are numerous bench scale items of equipment for sol-gel processing, moving bed materials snythesis, etc.
l 3.5.4 Special Projects Development Group The Special Projects development group provides materials selection, fabrication development and product characterization of all inorganic, non-metallic matericls in the nuclear fuel cycle; and all materials associated with nuclear scrap recovery, effluent treatment and non-aqueous, non-gaseous coolants.
)
1 i
License No.
SNM-778 Docket No.70-824 Date July, 1981 Page 3-14 Amendment No.
Revision No.
14 O
Babcock & Wilcox
I d
FIGURE 3-6 3
FAOILITIEC, V.'0FK ORD?.110R.*.1 4
TO Piant Eng;r.eering Dute Frem.
Section: _.-
Sigred:
Section Mgr :
Date:
I Date Required.
Choice No :
(Leber)
(Moterict)
DESCRIPTION OF WORK TO BE DONE i
I 4
l l
t l
i' 1
?
I 1
r 1
i A
I i
i I
t, i
I e,
i SIGN ATUR E RE QUleEO-tadustrial Safety Of ficer i
t l
f Health Physics:
Facility Supervisor:
j i
seece e....
. Line re, piene cas....rias vs. o.i,
.s i
, O rder Received Dore:
Signed-Plean ed Stortin g D ate.
Planne d Completion D ate:
Order Completeo:
Work Order Number Date-Sig n atu r e:
P
.. J.
,.-----,.-....-,.s.---,--~.-..--,..a...,-,-
L c
License N o.
SNM-778 Docket N o.70-824 Date O
A mend ment N o.
O Revision N o.
14 Page 3-33 o
l Babcock & Wilcox
-W C a2 Figure 3-7.
Building A Ventilation 0-D
!E Ez
??
u, E'~
5 A
a' !?
- i e E. %
Ez 5 N
,e 5
_P.A._
ROOF n
o
/
a tn 0".
Cf 4
O f -MOTORIZED DAMPER AHU sW( BLOWER O - FILTER BOX o
g PA - ABSOLUTE FILTER 88 L PR-AC AC-AIR CONDITIONING ACU-AIR HANDLING UNIT w
9 Bay-2 m
k.
5
-_b W
M L*
M
-._a O
O e
()
By anology, values for Pu would be similar.
The effect of interspersed water moderation in a concrete reflected finite array of 850 gram U-235 units is shown in Table 4-3(a). These data show that maximum array multiplication occurs with almost no interspersed water.
Table 4-3(a). K gg for 6x6x6 Array of 850 g U-235 Units on 30-in. Centers e
(y = 18.14 cm, p = 0.85 g U/cc)
Volume Keff 2a
+
Fraction H90 1.00 0.833 0.010 0.15 0.826 0.010 0.10 0.851 0.010 0.07 0.868 0.009 0.05 0.907 0.011 0.03 0.924 0.009 0.02 0.931 0.009 0) 0.01 0.937 0.009
(
0.001 0.930 0.0,09 The effect of varying the number of 850 U-235 units in a concrete reflected array while maintaining a constant center-to-center spacing with void between them is shown in Table 4-3(b).
Interpolating between the heterogeneous values for the 24 inch center-to-center system predicts a K
2a for 40 units of 0.936 ! 0.12 whereas the 36 inch spacing system g
has about 512 units for the same K ff.
License No.
SNM-778 Docket No.70-824 Date sonen-hor 1979 Page 4-9
[
Amendment No.
Revision No.
1
\\,i Babcock a.Wilcox
Table 4-3(b). K,gf for Arrays of 850 gram U-235 Units on 24 and 36 inch Centers Array Number eff Center-to-Center Size of Units 2a Spacing, in.
4x3x3 36 0.910 i 0.010 24 4x3x3 36 0.929 2 0.013*
24 4x4x3 48 0.932 0.010 24 4x4x3 48 0.947 0.011*
24 4x4x4 64
- 0. !'49 0.011 24 4x3x3 36 0.807 0.011 36 4x4x4 64 0.845 0.011 36 5x5x5 125 0.860 0.010 36 6x6x6 216 0.881 0.009 36 7x7x7 343 0.912 0.009 36 8x8x8 512 0.938 0.008 36 9x9x9 729 0.949 0.009 36 Assumes heterogeneous UO -water mixture.
2 From these data it is concluded that for 850 gras U-235 per unit an array on 24 inch centers would be safe for 40 units or less and on 36-inch centers an array would be safe with 90 units or less. A slight in-i crease in the array multiplication, on the order of 1%, may occur for low g
levels of interspersed water moderation. However, the safety of these arrays would still be maintained.
To avoid confusion and possible mistakes, additional pro-cedural controls are applied when low-enrichment limts 1
are used.
License No.
SNM-778 Docket No.70-824 D te July, 1981 p,g,4_10 Amendment No.
Revision No.
14 Babcock & Wilcox
()
These preclude enrichment combinations of below and above 4.0 wt% U-235.
(These are not necessarily unsafe -- no calculations were made and no such combinations are desired.)
4.
The unit and its limit (laboratory, furnace, transfer cart, etc.) are established by the facility su,ervisor, who author-izes posting the limit showing the maximum quantity of plutonium, U-233, and U-235 allowed. The fissile material content of the material transferred to or from a unit is established from process records, analyses, or previous analytical data. Only authorized users of SNM may transfer SNM between units and must do so only according to approved procedures. A board, sign, or other acceptable device is used to record the new balance and compares the balance with the unit lLait.
4.4.4 (A.9.5. 7) Building B 4.4.4.l (A.9.5.7.1) General -- The demonstration for units and the
/N array is identical to that of Building A (4.4.3.1 and 4.4.3.2).
(l 4.4.4.2 (A.9.5.7.2) Rot cell -- The demonstration for the units and array is identical to that of Building A.
The 5 units are isolated I
from all other arrays by a minimum of 2 feet of water or high density concrete.
4.4.4.3(A.9.5.7.3) Underwater Storage - Transfer Canal -- Underwater aluminum or stainless steel storage racks are constructed to ensure a 12-inch
~
edge-to-edge spacing of each unit. Units are limited to those'in A.9.5.6.2 excluding PWR fuel assemblies and, since they are separated by 12 inches of water, units are considered isolated. Therefore, any number of these units may be used at the LRC.
Racks and fixtures are constructed with sufficient integrity and sr.rength to withstand reasonable structural deformity, thereby providing the spacing previously outlined. Supervisory approval is required for removing or inserting any suberitical unit out of or into its storage rack.
l l
License No.
SNM-778 Docket No.70-824 Date
_ July, 1981 Page 4-11
/
Amendment No.
Revision No.
14 I
l Babcock & Wilcox
l i
There is no credible way in which water can be lost from the storage pool and transfer canal. However, assuming loss of water, stored units would drain and be unmoderated and suberitical.
4.4.4.4 (A.9.5.7.4) Underground Storage Tubes - Underground storage l
tubes are 5 inches in diameter, approximately 10 feet long, and on 17 inch centers (minimum) in a straight line. Material stored is first placed in a storage can with an inside .ameter of 4-1/2 inches. Maximum units demon-strated safe in Section 4.4.3.2 are stored one per tube. These are nuclearly isolated from each other by 12 inches of concrete (minimum). The average edge-to-edge separation approximates 13 inches of concrete.
4.4.4.5 (A.9.5.7.5) Destructive and Nondestructive Examination l
of PWR Fuel Assemblies 4.4.4.5.1 (A.9.5.7.5.1) General - The LRC will receive and examina by l both nondestructive and destructive examination PWR fuel assemblies.
Irradiated assemblies will have been subjected to a reactor environment with l
burnup to a maximum of 45,000 MWD /T.
From a nuclear criticality safety view-point, these assemblies are in their most reactive state when fresh or un-irradiated. Therefore, nuclear safety is demonstrated by appropriate
{
evaluations of the unirradiated assembly. The LRC curgent plans call for examination of B&W-manufactured fuel assemblies from B&W power reactors.
The current models of interest are designated as the Mark B and Mark C canless assembly. The Mark B assembly is described in the SNM license for B&W's Commercial Nuclear Fuel Plant (SNM License No. 1168, Docket 70-1201).
In 7.10 of Section III in SNM License 1168, the Keff of the unrodded ar.d fully moderated and reflected assembly is shown to be 0.92 at maximum l
enrichment. Maximum enrichment is defined as 4.0 percent nominal which could go to 4.05 percent in manufacturing. Table 4-4 shows a comparison of the Mark C and Mark B assemblies. The K gg of the Mark C assembly e
under the same conditions listed above has a value of 0.92.
The reactivity l
as well as the spectral and physics kinetics of these assemblies are i
essentially the same. All of the nuclear safety calculations shown in this section (4.4.4) were made with the Mark B assembly model (except License No.
SNM-778 Docket No.70-824 Date October 1979 4-12 Page
(
Amendment No.
Revision No.
2 l
Babcock & Wilcox
e Tables 4-4 and 4-5).
Results were obtained for a fully reflected infinite n
(v) array 12 inch edge-to-edge of maximum 1y enriched assemblies that were fully moderated, i.e., under water. The Mark B and C assemblies are to be disassembled in air only in an unirradiated state. The similarity in nuclear characteristics and the large decrease in reactivity in air-moderated assemblies assure nuclear safety during the disassembly operations.
Con-1 1
ditions given in A.9.5.7.5.1 are sufficient to ensure that these two assembly types are indeed those to be examined. A damaged assembly which is re-strained to 8.6 inches on a side will be no more reactive in air or water even if part of the fuel is missing; this will be demonstrated in Section 4.4.4.5,3.
Table 4-4.
Comparison of the Mark B and Mark C Fuel Assemblies Mark B Mark C Fuel assembly array 15 x 15 17 x 17 Fuel assembly dimensions, in.
8.45 x 8.45 8.536 x 8.536 Control rod tubes per assembly 16 24 n
Instrument tube per assembly 1
1 Fuel rods per assembly 208 264 Fuel rod pitch, in.
0.568 0.501 Fuel active height, in.
144 143 Pellet OD in.
0.370 0.324 Theoretical density, %
92.5 94.0 Enrichment, %
4.0 4.0 Fuel rod clad ID. in.
0.377 0.332 Fuel rod clad OD. in.
0.430 0.379 Fuel rod clad material Zr-4 Zr-4 V
/V in fuel rod cell 1.65 1.68 g,1 V
/V in assembly with water gg completely filling control rod and instrument cells 1.90 1.98 K,gg of one assembly in H O 0.92 0.92 2
License No.
SNM-77_8_
Docket No.70-824 Date October 1979 Page 4-13 Amendment No.
Revision No.
2 Babcock & Wilcox
f Since The Babcock & Wilcox Company is continuing to improve its assemblies and will supply reload fuel to reactors initially fueled by other taactor manufacturers, the LRC may destructively examine other types of assemblies. The conditions given for additional evaluation are adequate to ensure nuclear safety for different assemblies.
4.4.4.5.2(A.9.5.7.5.2) Receipt and Storage A.
Unirradiated Assemblies Unirradiated fuel assemblies may be stored in their shipping containers since their nuclear safety has been proven prior to their licensing. Assemblies that are unirradiated may also be stored in air if the distam ' between assemblies is no less than 21 by 38 inches.
(Refer to SNM-ll68, Tocket 70-1201, Ser. tion 3, page 185, dated 4/30/75). This distance assuris criticality safety for less than 100 assemblies of eit!.,e the Mark 8 and/or Mark C assembly types. This assures the safety of the maximum of four assemblies stored on site. Unirradiated assemblies may also be stored under water (hot cell pool, mock-up pool, or develc; ment test area pool).
(See Section 4. 4.4.5.2, B.)
Assemblies stored in air will be stored either:
1.
Horizontally -- on the floor or on tables constructed with sufficient integrity and strength to withstand reascnable structural deformity and assuring the above mentioned spacing.
2.
Vertically - in racks and fixtures constructed with sufficient l
integrity and strereth to withstand reasonable structural deformity and l
assuring the above mentioned spacing.
Supervisory approval is required to move any other fissile material into the area where the assemblies are stored. No more than four unirradiated assemblies may be stored at LRC at once. The limit of four l
assemblies is an arbitrary limit which the LRC imposes upon itself and does not affect nuclear safety.
1 License No.
SNM-776 Docket No.70-824 Date July, 1981 4-14 Page Amendment No.
Revision No. 14 Babcock & Wilcox
The safety of withdrawing an assembly and its associated rod storage p) position partially into the cell is demonstrated safe by comparison to a
(
series of KENO runs made for pool storage at a reactor site. The KENO physics code (a Monte Carlo code) is described in SNM License No. 1168, Appendix A to Sectirn III. pages 11 and 12.
To demonstrate the safety of flooding a reactor site storage pool filled with fresh Mark B fuel assemblies, an array of fuel assemblies 14 units wide, infinitely long, and reflected on the sides and bottom by concrete was calculated by KENO. Each assembly was spaced 1 foot from the other on the concrete re' lector, as appropriate.
Four cases at different degrees of pool flooding were evaluated and are de<.;ribed in Tab's 4-6.
Table 4-6.
Reactivity for an Infinite by 14-Unit Array of Fuel Assemblies Calculated K,ff Water Height Fully flooded 0.951 1 0.006 3/4 0.946 1 0.007 1/2 0.928 1 0.007 7 w,
(~_)
0 (dry) 0.50n 1 0.004 i
i l
l l
l License No.
SNM-778 Docket No.70-824 Date January,1980 Page 4-17 Amendment No.
Revision No.
5 l
Babcock & Wilcox
The confidence levels quoted above are one standard deviation.
K,ff for the fully flooded condition is higher than that calculated by PDQ because of simplifications made in running the cases. The series of runs were to demonstrate safety of a partially flooded pool, a much more restrictive condition than partial withdrawal into one cell. The similarity in the Mark B and Mark C nuclear characteristics and the simplifying assumptions assure these calculations are also valid for the Mark C assembly l
type.
4.4.4.5.4 (A.9.5.7.5.4) Assembly and Machine Shop and Development Test Areas Assemblies of either Mark B or C disassembled in air are far less reactive than the cases listed in Table 4-5.
Either assembly type may be disassembled in air. A safe reactivity level ("0.95) is assured provided the handling in Section A.9.5.7.5.4 is followed.The conditions stated in Section A.9.5.7.5.4 are based on KENO calculations that show:
1.
Two assemblies in air 21 inches or more apart are nuclearly safe.
2.
Fuel pins at a maximum enrichment when optimum 1y moderated are fully reflected in an infinite slab have a K,ff = 0.95 if the slab is no more than 4 inches thick.
3.
Fuel cods in any configuration or number, up to the number in the assembly, when limited to the confines of the assembly size are no more reactive than the intact assembly (Ref. Table 4-5).
4.4.4.5.5 (A.9.5.7.5.5) Hot Cell Operations - Work within the hot cell will, by and large, follow existing controls. One fuel rod will have a maximum of 2,230 grams of uranium or about 90 grams of U-235 at 4.0 wt%
nominal enrichment. Five rods represent, therefore, 450 grars of U-235 -
l considerablf less than the permitted unit of 850 grams U-235 at 4.0 wt%
enrichment or less. Two work stations are now authorized in Cell No. 1.
4-18 License No.
SNM-778 Docket No.70-824 Date Julv. 1981 Page Amendment No.
Revision No. 14 Babcock & Wilcox
pg 4.4.4.5.6 (A.9.5.7.5.6) Fuel Rod Dismantlement - Fuel rods of either assembly type may be dismantled. Dismantlement can be performed in any area for which present licensing conditions permit fuel handling.
In addition, mass control must be limited to 350 grams of U-235, proper spacing must be maintained, and approved procedures must be followed.
4.4.4.5.7(A.9.5.7.5.7) Shipment and Disposal - These conditions l
are consistent with the above demonstration and/or current limits.
4.4.5 (A.9.5.8) Building C l
4.4.5.l(A.9.5.8.1) General - Nuclear safety and nuclear isolation of l
the three described arrays are demonstrated as follows.
4.4.5.2(A.9.5.8.2)SNM Storage Vault - This vault is isolated by l
12-inch thick concrete walls of density >140 lb/cu f t.
A maximum of 40 units is permitted in the vault. Each unit stored therein is equal to or less than the maximum storage unit defined in Table IV, TID-7016, Revision 1 (as amended) except:
a.
The 350 gram U-235 limit with uncontrolled moderation, d
b.
The 850 gram U-235 limit for units with 4.0 wt% U-235 in U (also with uncontrolled moderation).,
c.
The 220 gram Pu limit with uncontrolled moderation.
All three of these were demonstrated as acceptable units in Paragraph 4.4.3.2 above. From Figure 22, TID-7016, Revision 1, 74 units in s cubic array are allowed for a 24-inch center-to-center, 8-inch edge-to-edge spacing, with full reflection. As previously demonstrated (4.4.3.2) the vault is limited to 40 units to accommodate low enriched units.
4.4.5.3(A.9.5.8.3) Process Area - The process area of Building C l
l is limited to 90 units on 36 inch centers, with at least 8 inches edge-to-edge between units. Each unit is limited to a mass value previously defined in 4.4.3.2 for Building A.
The allowable number of units according i
to Figure 22 of TID-7016, Revision 1, is about 190. The number of units has been reduced to 90 to permit the low enriched units.
l i
License No.
SNM-778 Docket No.70-824 Date October 1979 Page 4-19 Amendment No.
Revision No.
2 Babcock & Wilcox l
l
The values of all units in Building C are less than or equal to the value of the maximum storage unit defined in Table IV, TID-7016, Revision 1 (as amended), or they have been evaluated above in Section 4.4.3.2.
The allowable number of units on 36-inch centers is 90 units as evaluated in Section 4.4.3.2.
Administrative procedures for posting and controlling transfers of SNM to and from units are those described in 4.4.3.2.4.
4.4.6 (A.9.5.9)0utside Storage 4.4.6.1(A.9.5.9.1) General - The underground storage and shipments are nuclearly isolated by distance or matter.
4.4.6.2(A.9.5.9.2) Underground Storage - The underground storage tubes are 5 inches in diameter, approximately 20 feet long and 20-inch centers. Maximum units demonstrated safe in Section 4.4.3.2 are stored, one per tube. These are neutronically isolated from each other by 15 inches of concrete.
4.4.7 (A.9.5.10) Dry Waste Nuclear cri'icality safety of dry waste is ensured by maintaining the concentration of SNM to a value much less than an ever safe concentra-tion.
Forty-five grams of SNM in a 55-gallon drum yields a concentration of less than 0.25 g/ liter. These low concentrations are guaranteed by the nature of the material being stored which is contaminated laboratory waste'.
The nature of the waste as borne out by more than 20 years of experience will maintain an approximate uniform dispersion within the container. Dry vaste containers are stored in the radioactive waste building after gnema scanning to ensure that the maximum SNM is not exceeded. There is therefore no requirement in the number or arrangement of containers within the radio-active waste building. One dimensional transport caleviations show that, at a U-235 concentration of 0.25 g/ liter with optimum water moderation, a 5
fully concrete reflected sphere having the same volume as 8 x 10 55-gallon drums has a neutron multiplication of s 0.95.
Therefore, the 45 grans of U-235 per drum limit is safe in that the maximum number of drums on site can-5 not credibly exceed 8 x 10,
License No.
SNM-778 Docket No.70-824 Date July, 1981 Page 4-20 Amendment No.
Revision No. 14 O
Babcock & Wilcox
/
4.5 RADIATION SAFETY
\\~sY The responsibility for overall safety at the LRC is vested in the facility supervisors. The supervisor of Health and Safety is available to the facility supervisors for expertise in the fiel* of radiation safety.
The supervisor of Health and Safety reports to ths
' nager, Facilities Department, with direct access to the Director, L.C, when necessary. The Health and Safety group is comprised of the health physics and industrial safety staff.
4.5.1 Responsibilities The health physics staff is presently constituted as follows:
e A certified health physicsist e A health physics engineer Three health physics technicians e
The health physics staff is responsible for:
1.
General surveillance of all radiation activities.
l 2.
Distributing and processing personnel monitoring
()s equipment, maintaining individual exposure records,
(,
notifying supervisors of exposures greater than the permissible levels, and recommending appropriate restrictions.
3.
Leak testing radioactive sources.
4.
Supervising shipping and receiving of radioactive material.
I 5.
Supervising and coordinating the waste disposal l
i program.
6.
Assisting in personnel, equipment and facility decontamination.
l 7.
Radiation safety training.
8.
Providing expertise in all aspects of radiation protection.
9.
Generating or acquiring, maintaining and appropriately distributing all records and reports required by regu-lations or procedures that pertain to the staff's scope of interest.
License No.
SNM-778 Docket No.70-824 Date July, 1981 4-21 p,g, s
l
(
)
Amendment No.
Revision No. 14 Babcock & Wilcox
10.
Implementing the LRC's respiratory protection program.
4.5.2 General Surveillance General surveillance includes smear surveying, air sampling, water sampling, environmental sampling, effluent air monitoring, effluent liquid monitoring, and direct radiation surveys.
4.5.2.1 Smear Surveying - Smear surveying is performed in all areas which, in the judgment of the health physicist, have a potential for surface contamination. The frequency of smear sampling will vary depending on the potential for contamination, the previous experience with the area, and the need for keeping the area free of contamination. Schedules and areas listed in this section for discussion are subject to change.
l 4.5.2.1.1 Smear samples are taken with small absorbent, filter papers. The smear paper is moved across an area 2
of approximately 100 cm using about 5 pound pressure.
The smear may be counted with a portable gas flow proportional counter which is capable of detecting alpha or beta activity. Appropriate factors are applied to obtain results in disintegrations / minute.
Large area smears are taken of the reactor areas and main hallway in Building A, the hot cell operations areas, the change room and main hallways in Building B, and the laboratory areas of Building C.
Records of smear results and actions takan in response to these results are retained by the health physics group.
4.5.2.2 Air Sampling and Monitoring l
4.5.2.2.1 Air Monitoring Program - Air monitoring can be separated l
into two types: fixed continuously indicating and portable continuously indicating. The type or types to be used are at the discretion of health physics personnel.
1.
Fixed Continuously Indicating
(
l This type of air monitor, with a preset alarm level, is normally placed in each radioactive material handling area where, in the opinion of the health physicist, there is a potential for airborne activity.
License No.
SNM-778 Docket No.70-824 Date July, 1981 Page 4-22 Amendment No.
Revision No. 14 O
Babcock a Wilcox
sw
(#)
It is located where, in the judgment cf the health physicist, it will provide the most adequate service.
The alarm setting for a plutonium area is about the equivalent of one 40-hour work week at 1 MPC and cannot be set lower due to radon levels.
Alarm i
levels in other areas are set in accordance with the particular operation, the material being handled, and its potential hazard. The actual levels are then set as low as possible commensurate with local radiation levels.
2.
Portable Continuously Indicating This type of air monitor is placed as close as practicable to a particular operation cnd usually has the same alarm-level setting as the fixed continuously indicating monitor.
4.5.2.2.2 Air Sampling Program - Air sampling can be separated into two categories: fixed and portable.
Selection of the category is at the discretion of health physics personnel.
1.
Fixed A central racuum system with up to 100 sampling points O
has been insto led in Building C.
The sampling points
'b are located as close as practicable to operator stations to permit continuous sampling of breathing zones. Although these samples'are usally checked weekly, the frequency may vary as the situation dictates. Samples are usually counted after waiting for radon daughters to decay, but if a particular l
operation is in question, the samples may be counted after a shorter period, and an appropriate radon decay factor applied.
2.
Portable Where the use of a fixed air samp1tng h not practical, a portable sampler is used. This sampler is also used throughout the LRC to monitor special situations.
The samples are checked and evaluated like fixed air samples.
Air samples are counted on a low background proportional counting system. Appropriate factors for background activity and detector efficiency are applied to give results in disintegrations per minute. These results are License No.
SNM-778 Docket No.70-824 Date July, 1981 Page 4-23 j
Amendment No.
Revision No.
14 Babcock & Wilcox
than divid d by the total air flow volume of th2 system to obtain activity par unit volum3 (pCi/ml) which can be readily comparsd to MPC valurs fo:
particular isotopes.
The true flow rate is determined by the following method:
Initial flow rate + Final flow rate =
pg 2
4.5.2.3 Liquid Sampling - Liquid sampling consists of liquid waste tank samples, Lynchburg Pool Reactor (LPR) samples, hot cell storage pool, the experimental pool and pool gate samples. These liquid samples are taken at intervals determined by the health physicist.
Liquid waste tanks are sampled quarterly, prior to release to the Naval Nuclear Fuel Division's waste treatment system or whenever an unknown quantity of fissile material has been released to the tanks. Prior to sampling, the tank contents are thoroughly mixed and sampled. Two of the waste tanks are treated with strontium carbonate to precipitate radio-to settle strontium and they are mixed, but time is allowed for the SrCo3 before the sample is taken.
The LPR primary water is sampled each week during which the reactor has operated. The primary and secondary coolant is sampled monthly regard-less of operations. These samples are taken by the operations staff.
Sampling intervals of the other pools are dependent on use and are estab-lished by health physics.
Known amounts of sample liquid are extracted, evaporated to dryness and counted. Results obtainsd are recorded in activity per unit volume.
These records are retained by the health physics group.
4.5.2.4 Environmental Sampling - Environmental sampling of areas sur-rounding the LRC is performed to evaluate changes in levels of radioactivity of air, water and vegetation. As a minimum the program includes:
0 One continuous on-site background air sample.
9 Monthly water samples of the James River up and down stream of the liquid discharge point.
9 Continuous sampling of rain water on site.
9 Quarterly samples of river silt and near-river vegetation.
License No.
SNM-778 Docket No.70-824 Date November 1979 Page 4-24 Amendment No.
Revision No.
1 O
Babcock & Wilcox
l
(
)
Environmental samples are collected by LRC personnel. Analysis is performed at the LRC or by a commercial firm.
4.5.2.5 Effluent Air Monitoring - Potentially contaminated air from hoods, hot cells, and glove boxes is discharged up the 50 meter stack. This stack is monitored. Particulates are removed by single stage HEPA filters.
The more hazardous operations require filtration through two stages of HEPA filters. Generally exhaust air containing beta-gamma activity is passed through a single stage filter, whereas that air from Pu glove boxes operations require filtration through two stages of HEPA filters.
The stack is powered normally by an electric eator which operates off site power. Emergency power is supplied by an internal combustion engine.
It is coupled to the blower shaft through a centrifugal clucch.
On loss of site power, the engine starts automatically and automatically takes the load as it comes up to speed.
The stack air sample is drawn from a sampling head located in the stack approximately 25 feet from the base. Sample air passes through a fixed filter paper, an activated charcoal cartridge, the chamber of the OC/
gas detector, a vacuum pump and returns to the stack. The fixed filter paper is monitored continuously for alpha and beta activity by a gas flow proportional monitor. The charcoal cartridge is replaced on a regular basis j
and analyzed by gamma spectroscopy. The radioactive gas monitor utilizes a halogen-quenched GM tube which monitors continuously. The continuous monitors are provided with alarm setpoints which are determined by health physics. The alarms are connected to an alarm panel loc'ated in Building B.
~
~
~
'~
Since the stack dilution factor is expected to be greater than 10,000 for at least 50 percent of the time, it is proposed that an allowable dilution factor of 3000 be applied to the discharge stream averaged over a period of 1 year, and a dilution factor of 30,000 allowed for favorable meteorological periods averaged over a 4-hour period.
The stack monitor ilow rate is maintained at a minimum of 2 cfm.
1 l
For gaseous activity, the isotopic activity (pCi/ml) is directly proportional to the instrument reading (counts / minute). For particulate activity, which l
p 4-25 License No.
SNM-778 Docket No.70-824 Date ~ September 1979 Page Amendment No.
_ Revision No. 1 l
l i
Babcock & Wilcox
is " trapped" by the filter paper, the isotopic activity is proportional to the buildup of activity per unit time.
For existing monitors operating at a flow rate of 6 cfm, the counter efficiencies are as follows:
85Kr 2.9 x 10-9 uCi/ml counts / min Beta particles 1x 10-10 pCi/ml = 800 counts / min increase hour Exhaust air from areas thich have a low potential for airborne activity may be exhausted directly to the roof if approved by the Safety Review l
Committee. Room air in areas equipped with continuous air monitors may be exhausted to the roof through HEPA filters, if the concentration of airborne radioactive material is below the appropriate MPC for an unrestricted area, and if approved by the Safety Review Committee.
Exhaust systems that cannot be practically discharged tc the atmo-sphere through the main stack, and where there exists a reasorable probability that the discharges could exceed 10 percent of the applicable MPO for an unrestricted area will be monitored by taking a contincus particulate sample.
Specific exemption is requested to allow the use of a dilution factor for discharge of radioactive material through a 50 meter stack as calculated by Gifford's solution to the ditfusion equation. Pasquill's turbulance Type D will be used to calculate an average dilution factor for the year, and Pasquill's turbulance Type C will be used to calculate the average dilution factor during a planned release under favorable meteorological conditions.
4.5.2.6 Effluent Liquid Monitoring - An underground tank farm (shown schematically in Figure 4-3) collects liquid wastes from all radio-active material handling areas (Figs. 4-4 and 4-5).
Hood, floer, and sink drains are normally connected to this system. Wastes from each area drain into specified tanks, enabling separation of wastes at their scurce. For example, drains from the Building C laboratory area are expected to contain License No.
SNM-778 Docket No.70-824 Date July, 1981 page 4-26 Amendment No.
Revision No.
14 Babcock & Wilcox
-/
alpha activity, and empty into a specific waste tank.
When tanks are full, the contents are analyzed, and if the analysis indicates concentrations within the prescribed limits for unrestricted areas in 10 CFR 20, the liquid is pumped to the James River via the NNFD waste disposal system. In the event the activity level of a weste tank would not permit release, dilution techniques or waste treatment methods are applied to reduce the activity to acceptable levels.
The dilution technique involves the addition of clean (uncontaminated) process water to the liquid waste, either as it is being released or while the liquid is scili in the tank.
One technique of waste treatment that has been used at LRC for several years involves the coprecipitation of radiostrontium with large amounts of nonradioactive strontium as strontium carbonate. Advantages to this technique include:
High decontamination for radiostrontium.
e High decontamination for other easily hydrolyzed e
elements.
7,
(_,)
e No introduction of undesirable additives, as carbonate is naturally present in bodies of water exposed to the atmosphere.
Isotopic dilution of radionuclides.
e 90Sr and 50 for 60Co have Decontamination factors of approximately 200 for been demonstrated by the addition of 250 pounds of SrCo3 and 1000 pounds of Na2CO3 to a waste tank containing approximately 3000 gallons of liquid.
The decontamination factor for radiostrontium was determined by using 85Sr as a tracer.
Periodically, it will be necessary to remove sludge from the bottom
(
of the tank, at which time the SrC03 carrier will be replaced.
t l
1 I
l t
,f-~y License No.
SNM-778 Docket No.70-824 Date September 1979 4-27 p,g,
'w/'I
\\
Amendment No.
Revision No.
1
(
l l
l Babcock & Wilcox
Nuclear crit'cality control for the liquid waste tanks is ensured lh by maintaining concentrations of fissile material well below critically safe levels. Waste tanks, which receive liquid waste from areas that handle significant quantities of SNM, are visually inspected each time they are emptied tc ensure that there is no accumulation of sludge in the tank.
Disposal o' contaminated liquids with a concentratico greater than 0.01 gram fissile ma :erf al/11ter in the liquid waste system is prohibited.
l A summary of total activity released via the liquid waste disposal system for the past several years is presented in Table 4-7.
Table 4-7.
Liquid Waste Released to James River, microcories 1975 1976 1977 1978 July-Dec Jan-June July-Dec Jan-June July-Dec Jan-June 14.8 Cr-51 Mn-51 21.0 253.3 18.1 5.6 7.6 Co-58 82.5 1,043 26.8 6.2 2.7 0.19 Co-57
{l Co-60 95.5 82.0 1,111 99.7 72.1 151.5 Fe-59 7.6 Sn-65 Sr-90 20.25 8.7 80.4 13.3 42.'
8.6 Y-90 21.25 8.7 71.3 13.3 42.1 8,6 Nb-95, Zr-95 8
0.3 12 0.7 28.4 Ru-106 Cs-134 407.3 67.5 25.4 5.7 58.5 96 Cs-137 5138.8 1,067 348.6 73.3 1,146 1,990 Ba-140, La-14G 2.3 Ce-144 Gross beta 82.6 15.0 159.1 260.2 72.1 248 Uranium 44.3 15.6 57 37.2 188 80.5 July, 1981 4-28 License No.
SNM-778 Docket No.70-824 Date Page Amendment No.
Revision No. 14 Babcock & Wilcox
)
,m
-(,)
4.5.2.7 Direct Radiation Surveys - Radiation surveys are made, at the direction of the health physicist, in order to assess radiation hazards produced by radioactive materials. Sufficient surveys are made in order to cemply with 10 CFR 20.
Appropriate radiation survey instruments are chosin for surveys based on the types and icvels of radiation expected in the area. The fol-lowing types of instruments are available.
Ionization Chambers These instruments measure beta and gamma radiation with a seneitivity of a few mR/h to several R/h.
h GM Meters These instruments utilize a Geiger-Muller tube which detects beta and gnmmn radiaticn, with a sensitivity of approximately 0.05 mR/h (normal background) to several R/h.
Neutron Rem Cnenter This instrument measures neutron radiation with a sensitivity of less than 1 mrem /hr to several Rem /h.
/
_s x t
1
\\#
Survey resulte are recorded and the record retained by health nhssics personnel.
4.5.3 Personnel Monitoring and Control of Radiation Exposure Health physics is responsible for distributing personnel monitoring equipment to employees and visitors. Monitoring devices are processed or read either by health physics personnel or a commercial firm.
Although it is the responsibility of each employee to keep track of his exposure and be aware of the applicable limits, health physics also monitors expasure records, and reports exposures greater than LRC administra-tive limits to the employee's supervisor.
l Health physics also advises employees on methods to maintain 1
exposures as low as reasonably achievable.
l License No.
SNM-778 Docket No.70-824 Date July, 1981 4-29 Page Amendment No.
Revision No.
14 Babcock a Wilcox
7 4.5.3.1 Personnel Dosimetry for LRC Personnel - All LRC personnel will be monitored for radiation exposura while at the LRC site. This will be accomplished in two ways:
1.
All permanent LRC personnel will be issued a LD dosimeter which will be attached to the ID badge.
2.
LRC personnel will be categorized as radiation and nonradiation workers.
In addition to the TLD worn on the ID badge, the radiation workers will be issued either a film badge or TLD badge. Two indirect-reading pocket dosimeters will also be issued when film is used. The required dosimetry will be worn by the employee at all times while at the LRC site.
4.5.3.1.1 TLD Dosimeters Attached to ID Badge - Each employee will be issued a thermoluminescents dadmeter (TLD). This dosimeter will be attached to the employee's ID badge. The employee will be required to either wear the badge in an exposed location, such as the shirt collar, or to keep the badge in his wallet.
In both cases, the employee will have the dosimeter on his person at all times he is on the LRC site. Since secretaries do not normally carry a wallet separate from their purse, they will be allowed to leave this dosimeter in their purse, at their desk, if they wish. They may go to all areas on the site except any area marked as a radiation area.
All persons wearing only a single TLD attached to ID badge will be permitted tu visit and work in radiation or radioactive materials areas provided l
their estimated whole body dose does not exceed 100 millirems for the year.
No employee wearing only a single TLD attached to ID badge will be permitted in a high radiation area or in an airborne activity area.
4.5.3.1.2 Radiation Workers - Radiation workers, or employees whose projected or past yearly whole body dose is 100 millirems or more, will wear a film badge and two indirect-reading pocket dosimeters * (two TLD's may be used in place of these). Tais dosimetry will be carried at all times while on the Mt. Athos site. When not in use, the film badge and pocket dosimeters are required to be left in the film badge rack at the LRC. The License No.
SNM-778 Docket No.70-824 Date September 1979 Page 4-30 Amendment No.
Revision No.
1 O
Babcock siWilcox
(')'N TLD attached to the ID badge does not have to be worn in close proximity with
\\
the film badge, however, it murt be carried on the person of the worker.
Personnel will wear film badge or TLD badge when they visit or work at another installation (including the CNFP and 'the NNFD) for periods under 30 days, where they could be exposed to radiation. If visiting or working exceeds 30 days or extends over the first of the month, personnel must contact health physics for additional instruction concerning the changing of the film badge or TLD badge.
4.5.3.1.3 Types of Monitors - Film badges monitor external exposure to gamma, beta and in some cases neutron radiations. These badges are pro-cessed (read, recorded, and reported) by a commercial concern every month.
Beta and gamma range is from 15 to 500,000 millirems.
TLD's measure gamma exposure; the range is from 10 millirads to 10,000 Rads.
They will be read at yearly intervals or less.
Indirect-reading dosimeters measure external exposure to gamma radiation (range is 0 to 200 millirads) and are read and recorded daily.
If a TLD is used in place of dosimeters, it will be read on a frequency
[O h
determined by the health physicist. Persons handling radioactive sources shall wear finger rings also, when deemed advisable by health physics; these rings monitor gamma and beta exposure to the hands (range is from 50 millirads to greater than 10,000 Rads).
Direct-reading pocket dosimeters will be worn also, when deemed advisable by health physics. These dosimeters measure accumulated expo-sure when personnel are working in variable radiation fields. They mea-sure external exposure to gamma radiation from 0 to 200 millirads, or from 0 to 500 millirads.
All new employees and all new persons monitored as employees must b( initially admitted using Form LRC-129 and must receive proper indoctrination.
- The determination and estimation of 100 millirems exposure will be made
_He will take into account the job, assignment, by the health physicist.
past exposure records, etc., in his decision._
License No.
SNM-778 Docket No.70-824 Date September 1979 pag, 4-31
,_(,)
i Amendment No.
Revision No.
1 Babcock & Wilcox
The limits of exposure (see 10 CFR, Part 20) are as follows:
1.
Whole body, 1.25 Rem / calendar quarter.
2.
Skin, 7.5 Rem / calendar quarter.
3.
Hands, forearms, feet and ankles, 18-3/4 Rem / calendar quarter.
4.
Whole body, 300 mrem / week; the long-term exposure is con-l trolled by the supervisor.
During extreme emergencies, section and laboratory managers are authorized to exposure their personnel to 3.0 Rem / calendar quarter of whole
)
body radiation. At other times, the LRC Director may authorize whole body exposure of 3.0 Rem / calendar quarter.
1 4.5.3.2 Personnel Dosimetry for Non-LRC Personnel - No visitor may receive a radiation exposure exceeding 10 millirems in one week unless he is working under a radiation work permit * (Fig. 4-6).
I Certain non-LRC employees may be monitored and badged as cmployees; currently these include selected personnel assigned to the Facilities Department. If there is a question, health physics will decide whether a person will be monitored as a visitor or as an LRC cmployee. Non-LRC employees will be badged and monitored to comply with the following sections.
4.5.3.2.1 Category-A Visitors - A person visiting the LRC for one day or less and not covered by any of the following categories will sign in and sign out with the receptionist upon each visit. The receptionist will issue one film badge and one pocket dosimeter or two TLD's to each Category-A visitor. The visitor will leave the monitoring equipment with the receptionist upon signing out.
- This statement means that visitors have access to nonradiation areas at the site under the supervisiron of the LRC person they are visiting.
If their visit takes them into a radiation area, the LRC person being visited must make a stay-time calculation of the dose. The dose estimate must be less than 25 mrem / week, or the visitor must visit under a radiation work j
permit.
l l
License No.
SNM-778 Docket No.70-824 Date July, 1981 Page 4-32 Amendment No.
Revision Nc,.
14 l
l t
Babcock a.Wilcox
+.
()
3.
Rinse thoroughly.
If monitoring indicates, repeat up to three times.
(Do not wash often enough to cause skin irritation.)
4.
If this procedure fails to reduce levels to those specified in Table A-3 of Appendix A, contact health physics for further instructions.
4.5.7.2 Equipment and Facility Decontamination - Special care must be taken in decontaminating equipment and facilities where high levels of contamination exist. The possibility of personnel contamination, ingestion, or inhalation of radioactive materials may exist.
Where possible, decontamination should be performed using wet techniques to minimize the spread of contamination. Normal methods would include washing with soap and water or washing with special solutions such as commercially available decontamination agents. In certain cases, the health physicise may recommend the use of weak acid solutions or ultrasonic cleaning.
The permissible contamination allowed in " uncontrolled" and
" contamination" areas is specified in Table A-1 of Appendix A.
Maximum per-missible contamination levels for clothing and equipment are shown in
(')
Tables A-4 and A-5, respectively.
v 4.5.8 Radiation Safety Training Programs Radiation safety training at the LRC consists of three programs of various levels which are presented to LRC employees according to the type of work to be done by the employee. Also included is an annual re-training session for authorized users of radioactive material. Special training sessions may be conducted to indoctrinate personnel concerning special operationa or pertinent new regulatory requirements. All training programs are conducted by or under the supervision of the health physicist.
Reccrds pertaining to training programs will be retained by health physics personnel.
The formal training programs conducted at LRC follow.
License No.
SNM-778 Docket No.70-824 Date September 1979 Page 4-37 s,
\\s/
Amendment No.
Revision No.
1 Babcock & Wilcox
T l
\\
Program I - This course, presented to new employees shortly aftar joining LRC, provides an introduction to radiation theory (understandable to persons without technical education or experience) and a thorough coverage of radiological safety procedures and rules in effect at LRC.
Subj ects include types of radiation, effects of radiation on humans, exposure limits, a history of health physics, emergency and evacuation procedures, and basic health physics and personal hygiene. Completion of this course does not l
authorize an employee to handle radioactive material.
It is presented for general information. Therefore, no effectiveness evaluation is performed.
l l
Program II - This program, presented as a series of lectures and tests, is presented to new employees who will do a significant amount of I
work with radioactive materials. The content of the program is modified periodically to reflect changes in procedures, new regulations, etc. The effectiveness of this course is evaluated by written examination and by observations made during the monthly health physics audit. General course contents will include:
1.
Radioactivity a.
Types of radiation b.
Radioactive decay c.
Padiation dose and dose rates d.
Protection factors - time, distance, and shielding e.
Radiation effects on living systems f.
Radiation sickness g.
Effects of radiation exposure as compared with other common hazards 2.
Health Physics Instruments a.
Ionization chamber b.
GM counter c.
Alpha counter d.
Air monitors e.
Criticality alarm system f.
Emergancy equipment g.
Instructions in field use of instruments License No.
SNM-778 Docket No.70-824 Date July, 1981 Page 4-38 Amendment No.
Revision No. 14 O
Babcock & Wilcox L
l
[]
3.
Regulations and Procedures a.
Code of Federal Regulation b.
License requirement c.
Ishipment of radioactive materials d.
Waste disposal e.
Internal procedures Program III - This program is presented to technical and scientific personnel who are knowledgeable and experienced in work with radioactive materials.
The course content consists primarily of Parts 2 and 3 of Program II, with extension of this material into more advanced areas. The effectiveness of this course is evaluated during the monthly health physics audit by observing workers following proper procedures and safety practices.
Annual Retraining - Authorized users of radioactive material and radiation workers are required to attend the annual retraining session in I
order to remain qualified fo: such classification.
The annual retraining course consists of a review of new regulations and procedures, and a review of established procedures and practices which the health physicist feels need to re-emphasize. Evaluation of this training is made during the monthly health physics audits by observing workers fol-lowing proper procedures and safety practices.
4.5.9 Radiation Protection Consultation Health physics personnel are available, when called upon, to offer consultation in radiation protection and safe handling of radioactive materials. This consultation would normally be at the request of LRC personnel, however, requests from other B&W divisions or B&W customers are also occasionally received.
I l
License No.
SNM-778 Docket No.70-824_ Date July, 1981 Page 4-39 Acendment No.
Revision No.
14 I
l Babcock & Wilcox
f Consultation might include advice pertaining to one of the following categories: Proper shipment of radioactie materials, maintaining exposures as low as reasonably achievable, decontamination techniques, exposure limits, shielding design, or Federal and State Regulations. Health physics person-nel are also called upon to help formulate procedures or sections of pro-cedures which relate to radiological safety.
4.5.10 Documents and Records The health physics group generates and maintains records and prepares and distributes required reports in accordance with applicable LRC pro-cedure: and State and Federal Regulatory requirements. These records and reports pertain to radiological safety and personnel protection, and include personnel exposure records, survey results, radioactive shipment records, instrument calibration records, radioactive material release records, and documentation of unusual occurences which pertain to radiation safety.
The following records will be maintained for at least 2 years.
1.
Records showing the results of surveys made to evaluate the radiation hazard incident to the production, use, release, disposal or presence of radioactive materials or other l
sources of radiation, as required by 10 CFR 20.401 (b).
2.
Records used to prepare NRC-4 Form, " Occupational Radiation Exposure History", as required by 10 CFR 20.102, 3.
Records showing receipt, transfer, export, and disposal of byproduct material.
4.
Records showing results of leak tests as performed on sealed sources in accordance with license conditions.
5.
Records showing receipt, shipment, and disposal of special nuclear materials and source material.
6.
When personnel monitoring is required by 10 CFR 20.202, as specified in 10 CFR 20.401 (a), records showing the radiation exposure of each individual will be retained in accordance with 10 CFR 20.401 (c).
7.
Calibration records for instruments used to perform surveys as required by 10 CFR 20.401 (b).
License No.
SNM-778 Docket No.70-824 Date _ July, 1981 page 4-40 Amendment No.
Revision No.
14 O
Babcock & Wilcox
/,_
4.5.11 Respiratory Protection Program I
i
\\s /
The health physics group is responsible for the implementation of the respiratory protection program.
The primary objective of a respiratory protection program is to limit the inhalation of airborne radioactive or other nazardous materials. This objective is normally accomplished by the application of engineering controls, including the use of process, containment, and ventilation equipment. When such controls are not feasible or cannot be applied, respiratory protective devices must be used.
The program will include the following:
1.
Procedures governing the selection, fitting, and use of respirators.
2.
Procedure for training of users of respiratory protection.
3.
Procedure for respirator decontamination, maintenance, and storage.
4.
Medical surveillance for users of respiratory protection.
5.
Regular inspection and evaluatioa of the program to determine its continued effectiveness.
Q 4.5.12 Audit The health and safety supervisor or his designee perform audits monthly.
A written report is to be filed with the Director, LRC quarterly with a copy to the License Administrator. The audits are conducted in accordance with a written plan. An example of the contents of an audit plan is:
1.
Each month the following items are audited.
a.
Records of shipments and receipts of radioactive material are reviewed for completeness.
b.
LRC-229 " Facilities Work Order Formf are reviewed to ensure l
that sporopriate signatures have been e.ntered.
c.
Selected work areas are inspected to ensure proper posting, labeling and storage of radioactive material, and safety practices and procedures are being followed.
License No.
SNM-778 Docket No.
7u-824 Date July, 1981 Page 4-41 Amendment No.
Revision No.
14 Babcock & Wilcox i
f 2.
During the year the fa uowing items are covered:
a.
An evaluation of the respiratory rt:otection program.
b.
An evaluation of employee whole body exposure, c.
An evaluation of environmencal releases.
d.
An evaluation of bioassay results.
e.
An evaluation of airborne radioactivity.
f.
An evaluation of environmental monitoring.
g.
An audit of health physics records.
4.5.13 ALARA The managemt.nt at the Lynchburg Research Center is committed to the philosophy of maintaining radiation exposure to levels that are as low as rec.sonably achievable. This commitment is made known to all employees within the first month after they report to work. Employees wno's work assignment involve exposure to licensed material are provided with the initial training and annuai retraining that reinforces thfa commitment and provides the employee with the fundamental knowledge necessary to assist in implementing the ALARA princ.iple.
The inplementation of the radiation protection program is the responsibility of the health and safety supervisor. The qualifications of this position are presented in Section 2.
These qualifications ensure that the individual holding the position is eminently capable af dealing l
with potential problems encountered at the LRC. The supervisor has the authority to suspend operations until corrective action is taken if he observes practices which could result in a finite hazard.
The health and safety supervisor is responsible for reviewing personnel exaosures to assure that ALARA principles are being applied.
Table 4-8 is a summary of whole body radiation exposu.e for the calendar year 1977 and 1978.
It can be seen by this data that there is an increase in exposure dose in 1978 as compared with 1977. This is an example of a trend in exposures that resulted from an increase in the level of work I
in licensed activities rather than a reduction in the application of License No.
SNM-778 Docket No.70-824 Date July, 1981 Page 4-42 Amendment No.
Revision No.
14 O
Babcock & Wilcox
ALARA principles. This type of trend is not unusual in research and
,_()
development where the level and types of work will vary from year to year.
In this case the increased exposure is attributable to an increace in hot cell work and support activities from 5.75 manyears of effort in 1977 to 9.75 manyears in 1978 and an increase in critical experiment operations over the same period.
During this period no abnormal occurrences resulted in internal or external exposures.
Air sample data presented in Table 4-9 for plutonium and uranium areas historically have averaged less than 1% of MPC with an occasional sample indicating 2-3% of MPC. Bionssays including invivo for alpha emitters historically indicate zero exposure. Occasionally a sample will indicate detectable activity but analyses of subsequent samples have not confirmed an internal deposition.
Air sample data for. hot cell work has shown that some air samples including the respiratory protection factor utilize a significant portion of 40 MPC hours per work week. However, invivo counting indicates the presence
[V')
of only trace quantities of gamma emitters.
In most cases < 1% of a lung burden is indicated and up to 3% of a lung burden in a few cases.
An example of the application of ALARA principles is the improved method of fuel transfers into and out of the hot cell. Previously these transfers were made in air through a cell door resulting in an exposure of approximately 100 mR per transfer. By performing these transfers under water the exposures have been reduced to less than 10 mR per transfer.
Respiratogrprotection equipment for making entries into the hot l
cell has been filter masks. The installation of a supplied air respiratory system is planned.
Installation will get under way in the near future at a cost of approximately $70,000. This new system will result in an increase in the protection factor to workers of fron 50 for filter masks to 2000 for the supplied air system.
License No.
SNM-778 Docket No.70-824 Date July.1981 Page 4-43 Amendment No.
Revision No.
14 Babcock & Wilcox
(
4.6 FIRE SAFETY O
4.6.1 Fire Prevention 4.6.1.1 Housekeeping -- The facility supervisors are responsible for the day-to-day safety of operations at the LRC. This responsibility includes assuring that fire loading in their facility is maincained at the practicable minimum and work and storage areas are kept clean.
4.6.1.2 Area Operating Procedures - In exercising their review and approval responsibility for area operating procedures, the facility supervisors shall assure that new project and operations involve the least practicable amount of combustible and flammable materials. In exercising its review and approval responsibility, the Safety K2 view Committee also reviews new projects and area operating procedures for fire safety.
4.6.1.3 Flammable Liquids 4.6.1.3.1 General - Flammable liquids for use in the laboratories or work areas are stored in flammable liquid storage cabinets. Volatile liquids may be stored in glass or plastic containers outside a flammable liquid storage cabinet of not more than 1 liter capacity for laboratory l
I use.
Quantities exceeding 1 liter may be located outside a flammable lionid storage cabinet but only in safety cans.
4.6.1.3.2 Hot Cells and Glove Boxes - Hot cells and glove boxes are limited to quantities of volatile material which, when vaporized and mixed throughout the volume of the hot cell or glove box, would not result in the accumulation of an explosive mixture. The determination of the maximum permissible concentration is made by calculation. An example of the calculational method is presented below for Methyl Alcohol in Hot Cell No. 1 l
l License No.
SNM-778 Docket No.70-824 Date September 1979 Page 4-44 Amendment No.
Revision No.
1 i
E!abcock & Wilcox
1 i
Assumptions:
(G')
l.
Lower limit of flammability (% by volume) 6.7 2.
Volume of Hot Cell No. 1 54,374 liters 3.
A'verage hot cell temperature 3000k 4.
Average hot cell differential pressure 0.004 atm
(.*. pressure in hot cell = 0.996 atm) 5.
The ideal gas law applies n=
where:
n = number of moles of the material P = pressure in the hot cell (atm)
V = partial volume of the hot cell occupied by the material (liters)
T = temperature of hot cell air (Ok)
R = gas constant = 0.0821
" " (0.996 atm) x [(0.067) x (54374 liters)3 (0.0821) x (3000k) n = 147.3 moles 3(d Since CHg0 has 32 g/ mole and a density of 0.81 g/ml, the maximuu permissible volume would be:
147.3 moles x x
= 5.8 liters 0.
g 1000 ml Similar calculations are performed for other volatile materials in other hot cells and glove boxes. This method does not consider the air flow through l
the hot cell or glove box.
If a need arises for a larger volume than that calculated as above, the air flow may be used to demonstrate the safety.
l If the case of a combination of volatile materials is needed, a unity rule will be used for each hot cell or glove box as follows:
vol (volatile mat. A) vol (volatile mat. B) max. permissible vol A max. permissible vol B + ETC = <l.
l L
l July, 1981 l
License No.
SNM-778 Docket No.70-824 Date page 4-45 Amendment No.
Revision No.
14 l
l Babcock & Wilcox
Volatile material stored in safety cans within hot cells and g1 rue boxes are considered isolated and are therefore not included in the unity rule calculation. However, material removed from safety cans are included.
The facility supervisor verifies the correctness of the calculations and must authorize the introduction of flammable liquids into hot cells and glove boxes prior to the introduction.
4.6.1.4 Welding and Cutting -- Welding and cutting operations are performed in specially equipped, designated areas at the LRC. Such operations may be performed outside of buildings without restriction when carried out at distances greater than 35 feet from gas or gasoline storage areas. Welding l
and cutting operations are permitted in other than designated areas pursuant et a flame permit (Fig. 4-7).
4.6.1.5 Fuel Rod Sawing Fines -- Fines from fuel rod sawing operations I
in the hot cells are callected in a water-filled pan. When disposal of the fines is determined to be desirable, Met-L-X powder is added to the pan and mixed.
l The pan is then emptied into a primary waste receptacle and allowed to dry l
l at room temperature. When the water has evaporated, the container is sealed and is placed in a secondary container for eventual removal from the cell.
l k
When the primary container is sealed, the fines are considered isolated from the hot cell atmosphere and no longer poses a fire hazard. The number of these sealed containers that can be allowed to accumulate is determined by the quantity of SNM permitted in the cell.
{
4.6.1.6 Perchloric Acid - The use of perchloric acid is permitted in hoods specifically designed and built for this purpose. These hoods are equipped with wash down capability. No filters are installed in the off-gas line.
1 The use of perchloric ci' is controlled administrative 1y by 1
specifically approved Area Operating Procedures which specifies the restrictions placed on storage and use of perchloric acid and materials l
l that shall not be used or stored in perchloric acid hoods.
l
=
l l
License No.
SNM-778 Docket No.70-824 Date September 1979 Page 4-46 Amendment No.
Revision No.
1 l
lll>
Babcock & Wilcox
O O
O O
E n
e e
DIRECTOR s
s A.
e 1
3 2
l l
I I
?
Z FACILITIES PURCHASING ADMIi 5 VE VICES PERSOMEL o
it Manager Manager Manager Manager 2
I 1
m 0
HATERIALS & CHEMISTRY LABORATORY SENIOR SCIENTIST SYSTEMS DEVELOPMENT LABORATORY b
O Manager Manager
{
CERAMICS fiONDESTRUCTIVE METHODS N
3
& DIAGNOSTICS Z
Section Manager Z
0 o
o Section Manager EY s
u CHEHICAL & NUCLEAR g
ENGittEERING U
C v~
ao PROCESS CONTROL 5
~
Section Man:.p r g
& INSTRUMENTATION
's Section Manager
{ ')
O IRRADIATED HATERIALS g
a TECHfAOLOGY O
SYSTEMS DESIGM o
e tc)
& ENGINEERING e
O Section Ha'tager e
u X
5 Section Manager go 7
p NUCLEAR MATERIALS TECHNOLOGY L
~
C "J
N Section Mana;er H
O C
ca p
W Tip