Text

Onsite Radiological Contingency Plan for Rockwell Intl Operations
	ML20154C521
Person / Time
Site:	07000025
Issue date:	07/25/1988
From:	; ROCKWELL INTERNATIONAL CORP.
To:	;
Shared Package
ML20154A309	List: ML20154A306; ML20154C521;
References
	RI-RD88-206, NUDOCS 8809140396
	Download: ML20154C521 (108)
	v • d • e

....

Rl/RD 88 206

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ON SITE RADIOLOGICAL CONTINGENCY PLAN FOR ROCKWELL INTERNATIONAL OPERATIONS LICENSED UNDER SPECIAL NUCLEAR MATERIAL LICENSE NO.SNM 21 O

JULY 26,1988 i

i ROCkwellInternational O

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8809140396 000726 gDR ADOCK 07000025 PDR

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RI/RD88-206 i

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ON-SITE RADIOLOGICAL CONTINGENCY PLAN FOR ROCKWELL INTERNATIONAL OPERATIONS j.

LICENSED UNDER SPECIAL NUCLEAR MATERIAL LICENSE NO SNM-21 JULY 25, 1988 i

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CONTENTS

]

j Page 1.0 General Description of the Plant / Licensed Activity................

1 -1 1.1 Licensed Activity Deu ription................................

1 -1 1.2 Site and Facility Description................................

1-1 1.2.1 Site Description - Santa Susana Field Laboratory......

1-1 1.2.2 Facility Description - Rockwell International Hot Laboratory........................................

1-13 1.3 Process Descriptions.........................................

1-21 1.3.1 Typical Fuel Disassembly at the Hot Laboratory (8uilding 020)........................................

1-21 l

1.3.2 Typical Fuel Decladding at the Hot Laboratory (Building 020)........................................

1-24 2.0 Engineered Provisions f or Abnormal Operations.....................

2-1 2.1 Standards, Specifications, and Safeguards Features of Systems to Accommodate Abnormal Operations................

2-2 2.1.1 System to Detect Accidental Releases and Provide Alarms........................................

2-2 2.1.2 System Provided to Limit Release of Radioactive Materials.................................

2-8 2.1.3 Additional Features...................................

2-15 2.2 Adequacy of Engineered Provisions............................

2-17 3.0 Classes of Radiological Contingencies.............................

3-1 3.1 Classification System...................................

3-1 3.2 Event Classification and Response............................

3-2 3.3 Range of Postulated Accidents................................

3-3 4.0 Organization for Control of Radiological Contingencies............

4-1 4.1 Normal Plant Organization....................................

4-1 4.2 On-Site Radiological Contingency Response Organization.......

4-3 1

4.2.1 Direction and Coordination............................

4-4 4.2.2 Emergency Response Team Members Contingency Assignments...........................................

4-4 4.3 Off-Site Assistance to Facility..............................

4-6 4.3.1 Medical Treatment Facilities..........................

4-6 4.3.2 Medical Evacuation Service............................

4-7 1

4.3.3 Firefighting Backup...................................

4-7 RI/RD88-206 111

CON 1ENTS v

Page 4.3.4 ~ Police Assistance.....................................

4-7 4.4 Coordination with Participating Government Agencies..........

4-8 4.4.1 Federal Radiological Emergency Response Plan..........

4-8 4.4.2 00E Emergency Radiological Assistance Team............

4-13 4.4.3 California State Radiologic Health Section............

4-13 4.4.4 Los Angeles County Department of Health Services......

4-13 5.0 Radiological Contingency Measures.................................

5-1 5.1 Activation of Radiological Contingency Response Organization.................................................

5-1 5.2 Assessment Actions...........................................

5-1 5.3 Corrective Actions...........................................

5-2 5.4 Protective Actions...........................................

5-5 5.4.1 Personnel Evacuation from Site and Accountability.....

5-6 5.4.2 Use of Protective Equipment and Supplies..............

5-7 5.5 Exposure Control in Radiological Contingencies...............

5-8 i

5.5.1 Emergency Exposure Control Program....................

5-8 5.5.2 Decontamination of Personnel..........................

5-13 5.6 Medical Transportation.......................................

5-15 5.7 Medical Treatment............................................

5-15 5.7.1 Medical Triage........................................

5-16 6.0 Equipment and Facilities..........................................

6 -1 6.1 Control Points...............................................

6 -1 6.2 Comun i c a t i on s Eq u i pmen t.....................................

6 -1 6.3 Facility for Assessment Team.................................

6 -1 6.4 On-Site Medical Facilities...................................

6-3 6.5 Emergency Monitoring Equipment...............................

6-3 7.0 Maintenance of Radiological Contingency Preparedness Capability...

7 -1 l

7.1 Written Procedures...........................................

7 -1 7.2 Training.....................................................

7 -1 7.3 Tests and 0 rills.............................................

7-2 7.4 Review and Updating of the Plan and Procedures..............

7-3 k

Rl/R088-206 iv

(T CONTENTS Page 7.5 Maintenance and Inventory of Radiological Emergency Equipment, Instrumentation, and Supplies.....................

7-3 8.0 Records and Reports...............................................

8-1 8.1 Records of Incidents.........................................

8-1 8.2 Records of Prepa redness As suranc e............................

8-1 8.3 Reporting Arrangements.......................................

8-2 9.0 Recovery..........................................................

9-1 9.1 Reentry......................................................

9 -1 9.1.1 Radiological Criteria.................................

9-1 9.1.2 Industrial Safety Criteria............................

9-1 9.2 Plant Restoration............................................

9-2 9.3 R e s ump t i o n o f O p e ra t i o n s.....................................

9-2 Appendix A - Description of Implementing Procedures for Emergencies...

A-1 FIGl'RES b

N.

1 -1 Los Angeles Area..................................................

1-3 1-2 Santa Susana Field Laboratory Topographic Map.....................

1-6 1-3 Aerial Photograph of Santa Susana Site (June 1979)................

1-7 1-4 S S F L S e c t o r Ma p...................................................

1-12 1-5 Building 020 RIHL Floor Plan......................................

1-14 1-6 Building 020 Ventilation System Schematic.........................

1-18 1-7 Building 020 Ventilation System Flow Diagram......................

1-19 4-1 Rockwell International's Rocketdyne Division Organization.........

4-2 5-1 Atmospheric Dispersion for Limiting Working Inventories...........

5-3 l

6-1 SSFL Emergency Assembly Areas.....................................

6-2 TABLES 1 -1 Nuclear Materials Allowed Under Special Nuclear Materials License ShM-21....................................................

1-2 RI/R088-106 y

TABLES (Continued)

~~

Page 1-2 Population Distribution Surrounding SSFL (1970 Census)............

1-8 1-3 Population Distribution Surrounding SSFL (1980 Projection)........

1-9 1-4 Population Distribution Surrounding SSFL (1990 Projection)........

1-10 1-5 Population Distribution Surrounding SSFL (2000 Projection)........

1-11 2-1 Enclosure Requirements for Highly Radiotoxic Nuclides.............

2-13 2-2 Modification Factors for Enclosure Requirements...................

2-13 3-1 Decision Matrix - Radiological Contingency -

Off-Duty Hours....................................................

3-4 3-2 Classification Decision Matrix - Radiological incidents -

On-Duty Hours.....................................................

3-5 3-3 Action Guide to Radiological Events...............................

3-6 3-4 Exposures Resulting from Airborne Releases at Nearest Boundary....

3-7 5-1 Smear Test Surface Contamination Limits and Action Guides.........

5 -9 5-2 Radionuclide Grouping Criteria....................................

5-10 5-3 Survey Instrument Surface Contamination Limits....................

5-10 5-4 Acceptable Surface Contamination Levels...........................

5-14 5-5 A Partial List of Area Hospitals Capable of Radiation Accident Management...............................................

5-17 6-1 Personnel Monitoring Instruments..................................

6-4 6-2 Area Monitoring Instruments.......................................

6-5 6-3 Instruments for Measuring Releases to Environment.................

6-6 l

6-4 Effluent Monitoring Instruments...................................

6-7 6-5 Instruments for Monitoring Meteorological Conditions..............

6-7 7 -1 Emergency Supplies at Santa Susana Field Laboratory Site..........

7-4 1

1 I

Rl/R088-206 vi

/'l 1.0 GENERAL DESCRIPTION OF THE PLANT / LICENSED ACTIVITY V

1.1 LICENSED ACTIVITY DESCRIPTION Licensed activities are performed at Rocketdyne's Santa Susana Field Laboratory (SSFL) located in the Simi Hills about 29 miles northwest of down-town Los Angeles, California.

The Santa Susana licensed facility consists of Building 020, the Rockwell International Hot Laboratory.

The hot laboratory is used for the decladding and cleaning of irradiated reactor fuel and for hot-cell examination of ir-radiated nuclear fuels and reactor components.

The amounts of nuclear mate-i.als allowed for this facility under license SNM-21 are presented in Table 1-1.

A portion of the material in this document has been taken f rom the license renewal application document (ESG-82-33, revised June 5,1984, "Health p

and Safety Sections for Renewal Application of the Special Nuclear Materials License SNM-21, Docket 70-25').

More detailed information can be obtained with reference to that document.

The order of presentation here is consistent with the standard format for Radiological Contingency Plans.

1.2 SITE AND FACILITY DESCRIPTION 1.2.1 Site Description - Santa Susana Field Laboratory The Santa Susana Field Laboratory is the site of the only licensed facil-ity, Building 020, the Rockwell International Hot Laboratory (RIHL).

The SSFL is located in the southeastern portion of Ventura County, adja-cent to the Los Angeles County line.

The site is approximately 29 miles northwest of downtown los Angeles.

The location is shown in Figure 1-1.

Its distance f rom and directional relationship to various surrounding comunities is:

O Santa Susana 3 miles (4.8 km) north U

RI/R088-206 1-1

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-l Table 1-1.

Nuclear Materials Allowed Under Special Nuclear Materials License SNM-21 Byproduct. Source, Maximum Amount That Licensee May and/or Special Nuclear Chemical and/or Possess at Any One Time Under Material Physical Fons This License A.

Uranium enriched in the A.

Any enrichment or f6rm A.

5 kilograms U-235*

U-235 isotope except UF-6 3

B.

Pu (principally 239-Pu)

B.

Any B.

Maximum of 2.0 kilograms of s

total Pu*

(1) SSFL Site 7

(a) Hot Laboratory--Up to eu g

2.0 kilograms Pu in irradiated fuel, with e

less than 1.0 kilo-gram Pu in process C.

Pu (principally 239-Pu)

C.

Sealed sources (as C.

1.0 kilogram total Pu in any Pu-Be sources) building at the SSFL site in accordance with NRC approved radiation safety criteria

Enriched uranium and plutonium in combination not.to exceed 5.0 ef fective kg calculated by (kg U-235 + [2.5][kg Pu])

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LEGEND:

Figure 1-1 HOSPITALS 1.

Humana Hospital West Hills 2.

Nu-Med Regional Medical Center 3.

Northridge Park Foundation i

FIRE STATIONS (Los Angeles City) 4.

Fire Station 104 (Paramedic Ambulance) 5.

Fire Station 96 (Paramedic Engine)

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6.

Fire Station 106 r

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Fire Station 12 l

8.

Fire Station 107 l

(Los Angeles County) 9.

Chatsworth Unincorporated (Ventura County) i

10. Simi i

POLICE / SHERIFFS l

(Los Angeles City) i i

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Devonshire Division j

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12. West Valley Olvision i

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Van Nuys Division f

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(Ventura County) t 14.

East Valley Station i

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O RI/R088-206 1-4

Susana Knolls 3 miles (4.8 km) northeast

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Simi 5 miles (8.0 km) northwest Canoga Park 6 miles (9.6 km) east-southeast Chatsworth 6 miles (9.6 km) east-northeast Calabasas 7 miles (11.3 km) south Woodland Hills 7 miles (11.3 km) southeast Thousand Oaks 13.5 miles (22.7 km) southwest The site is on a plateau atop the Simi Hills and is relatively isolated from the surrounding communities.

Its isolation is further enhanced by its elevation, which places it 800 to 1,000 ft above the populated valley floors.

A topographic map of the site is presented in Figure 1-2.

An aerial photo of the Santa Susana site covering an area of about 4 square miles is presented in Figure 1-3.

The legend to Figure 1-1 also lists those hospitals, Ventura County Sherif f's of fices, and fire stations that are supportive in any emer-gency situation that may arise at the Santa Susana site.

The population distribution for the residential population out to 5 miles O.

by sector surrounding the site.is presented in Tables 1-2 through 1-5, based on the 1970 census data and projected into the future by decade for three decades

The accompanying sector map for this population distribution is pre-sented in Figure 1-4.

The daytime off-site population is not expected to be any different.

There are no nearby facilities such as hospitals, rest homes, or schools that would be a potential evacuation problem.

Access to the site is limited because of its remote location; there is access over paved roads both from the Simi Valley and the San Fernando Valley.

Adequate roads within the site assJrc movement of Rocketdyne/ emergency vehicles.

The locations of the hospitals, fire stations, and police and sheriff's offices that would be supportive in any emergency situation that might arise are indicated on the map (Figure 1-1) and the facilities are specifically identified in the legend to this map, f

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Growth rates f or this area are reported to range f rom 1.36 to 9.2%/yr with i

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an average of 5.2%/yr.

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Population Distribution Surrounding SSFL (1970 Census)

Distance (miles)

Sector 0-1/2 1/2-1 1-2 2-3 3-4 4-5 Total N-NNE O

O O

2,954 4.738 3,720 11,412 NNE-NE O

O 0

966 3,336 0

4,302 NE-ENE O

O O

0 1,937 1,254 3,191 ENE-E O

O O

0 5

396 401 E-ESE 0

0 0

2,335 2.335 ESE-SE O

O O

12 12 5,968

$,992 SE-SSE O

O O

23 5

1,020 1,048 l

SSE-S 0

0 0

S-SSW 0

0 0

SSW-SW 0

0 0

SW-WSW 0

0 0

WSW-W 0

0 0

W-WNW 0

0 0

WNW-NW 0

0 0

NW-NNW 0

0 0

NNW-N Q

Q Q

0 0

0 O

Total 0

0 0

3,955 10.033 14.693 28,681 O

RI/R088-206 1-8

O Table 1-3.

Population Distribution Surrounding SSFL (1980 Projection)

Distance (miles)

Sector 0-1/2 1/2-1 1-2 2-3 3-4 4-5 Total N-NNE 0

0 0

4,890 1,843 6,158 18,891 NNE-NE O

O O

1,599 5,522 0

7,121 NE-ENE 0.

0 0

0 3,207 2,076 5,283 ENE-E O

O-0 0

8 656-664 E-ESE O

O O

3,865~

3,865 ESE-SE O

O O

20 20 9,879 9,919 SE-SSE O

O O

38 8

1,689 1,735 SSE-S 0

0 0

S-SSW 0

0 0

SSW-SW 0

0 0

SW-WSW 0

0 0

WSW-W 0

0 0

W-WNW 0

0 0

WNW-NW 0

0 0

NW-NNW 0

0 0

NNW-N Q

Q Q

0 0

0 O

Total 0

0 0

6,547 16,608 24,323 47,478 O

RI/R088-206 1 -9

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Table 1-4.

Population Distribution Surrounding SSFL (1990 Projection)

Distance (miles)

Sector 0-1/2 1/2-1 1-2 2-3 3-4 4-5 Total N-NNE O

O O

8,095 12,984 10.195 31,274 NNE-NE 0

0 0

2,647 9,142 0

11,789 NE-ENE O

O O

O 5,308 3,437 8,745 ENE-E O

O 0

0 14 1,085 1,099 E-ESE O

O O

6.399 6,399 ESE-SE O

O O

33 33 16,355 16,421 SE-SSE O

O O

63 14 2.795 2,872 SSE-S 0

0 0

S-SSW 0

0 0

SSW-SW 0

0 0

SW-WSW 0

0 0

WSW-W 0

0 0

W-WNW 0

0 0

WNW-NW 0

0 0

j NW-NNW 0

0 0

NNW-N Q

Q Q

0 0

0 O

f Total 0

0 0

10,838 27,495 40,266 78,599 l

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Rl/RD88-206 1-10 I

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s Table 1-5.

Population Distribution Surrounding SSFL (2000 Projection)

Distance (miles)

Sector 0-1/2 1/2-1 1-2 2-3 3-4 4-5 Total N-NNE O

O O

13,402 21,495 16,877 51,774 NNE-NE O

O O

4,383 15,135 0

19,518 NE-ENE O

O O

O 8,788 5.689 14,477 ENE-E O

O O

0.

23 1,797_

1,820 E-ESE O

O O

10,593 10.593 ESE-SE O

O O

54 54 27,076 27,184 SE-SSE O

O O

104 23 4,628 4,755 SSE-S 0

0 0

0' S-SSW 0

0 0

SSW-SW 0

0 0

SW-WSW 0

0 0

WSW-W 0

0 0

W-WNW 0

0 0

WNW-NW 0

0 0

NW-NNW 0

0 0

NNW-N Q

Q Q

0 0

0

_0_

Total 0

0 0

17,943 45,518 66,660 130,121 O

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Rl/R088-206 1-12 1

1.2.2 Facility Description - Rockwell International Hot laboratory

]

1.2.2.1 Hot laboratory (Buildina 020) nE Operationt in Building 020 consist principally of th6 iisassembly, de-cladding, and examination of irradiated reactor fuel.

Hot-cell operations are conducted from the operating gallery where the in-cell equipment is remotely

]

operated.

The manipulators, analytical equipment, and controls for the vart-ous cell operations are located in this area.

The cells are serviced from the service gallery located to the rear of the hot cells.

Separating the cells f

and service gallery are the decontamination rooms, where equipment is decon-taminated prior to removal f rom the cells to the hot storage area.

The decon-tamination rooms also serve as contamination control areas between the cells and the service gallery. A hot storage area is provided for contaminated equipment. Also connected with the service gallery is a hot machine shop and a hot manipulator repair room for servicing low-level, radioactively contami-nated equipment.

In addition, controlled-environment gloveboxes are available 9r use with radioisotopes and low-level radiation operations. A machine shop 8

and mockup area allow fabrication and mockup of remotely operated equipment prior to installation in the cells.

The facility also includes change rooms, a photographic laboratory, and office areas. A plan view of the facility is given in Figure 1-5.

All cell doors are provided with inflatable seals to minimize leakage around the doors and between the areas.

Under normal operating pressures.

leakage around the doors is essentially prevented. With all doors shut and sealed, each cell and decontamination room can be operated independently.

For example, the oxygen content of the atmosphere in Cell 3 can be reduced by using a nitrogen purge, whereas normal cell ventilation is used in Cells 2 and 4; or Cells 2 and 4 ca" be entsred while an examination is croceeding in Cell 3.

T e

RI/RD88-206 1-13

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12-tosa urc t Figure 1-5.

Building 020 RIHL Floor Plan e

l

Thare is no pressurized water supply in the main cells.

A nitrogen atmosphere (less than 5% oxygen) is used to suppress, control, and extinguish fires.

Water to Cell 1 for the cleaning of metallographic supplies is pro-vided by a flexible tube through a shielded cell access port.

All pressure fittings and controls are mounted external to the cell.

The specific operations performed in the cells vary depending on current needs and changing program requirements; the work is principally research and development.

Thus, only general techniques and types of operations typical to hot-cell operation are listed.

The various capabilities for the cell block area are listed below.

1.

Cell 1 a.

Radioactive waste packaging b.

Solidification of liquid radioactive wastes l

2.

Cell 2 a.

Material testing Tensile testing Stress rupture and creep Testing Fatigue testing b.

NaK and sodium distillation c.

Visual examination d.

Density measurements e.

Dimensional measurements f.

Minor component disassembly g.

Fission gas collection h.

Isotope encapsulation 3.

Cell 3 a.

Olsassembly cell for irradiated materials b.

Sample preparation c.

Elox equipment d.

Cut-off wheel e.

Waste packaging O

RI/R088-206 1-15

(

f.

Visual examination

(\\

g.

Stereomicroscopic examination h.

Dimensional measurements 1.

Cask unloading 4.

Cell 4 a.

Hydrogen ar,aly *.

b.

Profilometer measurements c.

Annealing studies d.

Permeation testing e.

Major component disassembly and repair f.

Visual examination g.

Stereomicroscopic examination h.

Fuel canning i.

Dimensional measurements j.

Waste packaging k.

Cask unloading and loading 1.

Density measurements m.

'Gama spectrometry O

s/

n.

Autoradiography on capsule assemblies m

1.2.2.1.1 krj.te Confinement and Effluent Control in RIHL 1.2.2.1.1.1 Gaseous Effluent Control 3

The building ventilation systems were designed principally to control airborne contamination.

These systems direct the leakage of air from the out-side of the building into the main cells. The air flow is always from an area of lower contamination to an area of higher contamination within the building.

1.2.2.1.1.2 High-Volume Cell Ventilation Ventilation for the four hot cells and the decontamination room is pro-vided by a 12,500-cfm constant-volume bicwer.

A second identical blower is located in parallel and is automatically actuated in the event of failure of the primary blower. Both blowers are on an emergency power system.

Rl/R088-206 1-16

/

t f

The exhaust from the cells passes through prefilters 'ocated in each cell and then through filters in the basement.

The filters located in the basement are specified to be 99.95% ef f ective for particles >0.8 um in size, using a standard "cold" 005 test.

Cell ventilation is controlled by pressure instruments located in the operating gallery.

Under normal conditions, with all cell doors closed and sealed, the pressure in the cells is maintained at >0.6 in, of water, negative with respect to adjacent areas The pressure differential results in approxi-mately 100-cfm to 400-cfm leakage f rom the operating gallery into each cell.

Because the blower is a constant-volume blower, make-up air to attain the 12,500-cfm blower capacity is automatically added by a valve in the basement of the facility.

A schemat, diagram of the system is shown in Figure 1-6 and a flow diagram in Figure 1-7.

Sufficient ventilation system capacity is provided to create large flow rates into the cells when any cell door is opened or other breach made. When 3

a cell door is opened, about 4,000 ft of air is exhausted from the cell per minute.

This total corresponds to a flow rate of about 200 linear feet per minute through the opening into the cell, a rate adequate to prevent the re-lease of contamination f rom the cell into the adjacent decontamination reom.

An electrical interlock is provided so that a cell door cannot be opened when the high-volume exhaust ystem is not in operation.

1. 2. 2.1.1. 3 Low-Volume Cell Ventilation To supply an inert atmosphere in the cells for fire prevention and the protection of pyrophoric materials, a low-volume ventilation system is pro-vided.

This system can maintain a negative pressure in a cell of 0.05 to 0.10 in. of water with all doors closed and sealed.

This less-than-normal negative press'aro res.:lts in less air leakage into the cells and reduces the amount of inert gas make-up required to maintain the oxygen content below 5%.

Nitrogen is the only inert gas currently in use.

Operations are discontinued if the negative pressure is reduced to a value less t:;an 0.05 in, of water.

O Rl/R088-206 1-17

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RI/RD88-206 1-19

The supply of dry nitrogen to the cell is automatically shut off if the cell negative pressure drops to 0.05 in, of water with respect to adjacent areas.

This safeguard prevents evil pressurization.

1.2.2.1.1.4 General Posted-Area Ventilation The general posted-area ventilation blower provides exhaust for the hot change room, the hot side of the manipulator repair room, the hot laboratory, the hot shop, and the operating gallery.

Two identical 23,000-cfm constant-volume blowers are located in parallel to provide this exhaust. One blower is normally in operation, and the second or standby blower is automatically actu-ated if the first blower fails.

Only one of the two blowers is on the build-ing emergency power system.

However, unless the blower on the emergency power system is inoperative, the electrical circuit sequence ensures the provision of general building ventilation during a loss of line power.

The prefilters on all ventilation systems (indicated in Figure 1-6) rre changed frequently to extend the life of the high-efficiency filters, in-cell

.prefilters are changed during every cell clean-up or when the negative pres-sure in the cell is less than 0.5 in, of water.

1.2.2.1.1.5 Gaseous Effluent Control System The gaseous effluents from the hot laboratory are discharged to the atmo-3 sphere through a 73-f t-high stack at a nominal air flow rate of 23,000 f t /

3 min, which may be increased by 13,000 f t / min when the cell high-volume exhaust is operative.

The filtration efficiency is about 99.99% for the DOS aerosol used to test the filters.

Potential contaminants are the uranium isotopes U-234 U-235 U-236, U-238 Thorium, Cs-137, Sr-90, Y-90, Kr-85, 1-129, and Pm-147.

Stack sampling is performed to permit the measurement of particulate radioactive material discharged from the facility.

A gas N nitor is installed in the stack to measure radioactive gas discharges and to indicate accidental criticality of low-energy release occurring within a cell.

O R1/R080-206 1-20

1.2.2.1.1.6 Liauid Effluent Control Radioactively contaminated liquid wastes from all hot facility drains are collected in one 3,000-gal waste tank in the hold-up tank building located to the east of the hot laboratory at the perimeter fence line.

An attempt is made to absorb or solidify all highly contaminated waste in the cell at the time of generation.

Thus, most of the tank contents is generated during decontamination using water.

A weir box is used to catch large particles prior to their entering the hold-up tanks.

Nonradioactive liquid wastes f rom the hot laboratory are combined with waste streams from other facilities at SSFL as well as with waste water from other Rocketdyne facilities before heing treated and released to uncontrolled areas.

1.2.2.1.1.1 Solid Wastes Solid wastes contaminated with radioactivity are packaged in approved O

containers and shipped off site for commercial land burial.

I 1.3 PROCESS DESCRIPTIONS 1.3.1 Iypical Fuel Disassembiv at the Hot Laboratory TBuildina 020) 1.3.1.1 Process Description The following is a process description of the Fermi feel disassembly at the Rockwell International Hot Laboratory, which is typical of operations cur-rently planned for that facility.

The fuel assemblies consist of an outside wrapper tube, which is a square stainless steel tube measuring 2.646 in, on a side with a nominal wall thick-ness of 0.0096 in.

O RI/R088-206 1-21

The fuel region of the assemblies is made up of 140 round U-10 w/o Mo O

alloy pins initially containing uranium enriched to 25.34% U-235, and which now contain a maxiinum of 4,806 g of U-235 and a maximum of 23.7 g of Pu-239.

l The nominal overall length of the pins is 32.8 in., and the outside diameter is 0.158 in.

Each pin is clad with 0.005 in, of reactor-grade zirconium that is metallurgically bonded to the fuel alloy by coextrusion and closed at the top and bottom with zirconium end caps. The lower end of each pin is fastened to the support structure by anchor bars that are inserted through slots in the i

bottom end caps.

The upper ends ire free to accommodate changes in length i

resulting f rom temperature and growth due to radiation of fects.

4 The support structure of the fuel pins has ten restraining grids and

]

seven guiding grids.

The support structure holds 140 fuel pins on a square l

pitch of 0.199 in.

The four corner positions are steel tie rods.

The grids l

are made of 347 stainless steel strips 0.0115 in. th'ick by 0.50 in, wide, which are slotted and dimpled.

i incoming Fermi fuel assemblies are stored in below-grade vaults at the q,

Radioactive Materials Disposal Facility (RMDF), s DOE-owned license-exempt facility operated by Rockwell International at the Santa Susana Field Labora-tory (SSFL).

Fermt fuel assemblies are transferred f rom the RMDF in an ap-l i

proved on-site transfer cask to the RIHL.

They are transferred one at a time and placed in Cell 4.

Only one assembly will be in Cell 4 at any time; how-ever, up to 5,000 equivalent grams of the SNM may be in (iie f acility.

In

{

Cell 4, the end hart 1 ware is cut of f the 140-pin assembly, using a band saw l

[

equipped with a water drip system.

The water drip system prevents any spark-ing of uranium fuel which could, in turn, cause ignition of zirconium chips.

i The 140 fuel pins are weighed in four batches of 35 pins each (to accom-i modate the maximum weight limit of the existing in-cell balance).

After l

weighing, the four batches of 35 pins are placed into the inner shipping can-ister and the bottom end cap is press fit into place and crimped.

The inner shipping canister is placed into the transfer can, which is transferred to the RMDF in an on-site transfer cask for storage.

Scrap hardware is separated in-l to three classifications:

high, medium, and low.

The high-level scrap is l

i O RI/R088-206 1-22 1

5

packaged in aluminum cans and remotely welded and leak tested at the RIHL.

It 6O is then tr'ansferred from the RIHL to the RM0F in an on-site transfer cask and stored.

The medium-level scrap is packaged in drums that are shielded with carbon steel.

The drums are transferred from the RIHL to the RM0F and stored prior to shipment to a licensed burial site.

The low-level scrap is packaged in 00T-approved boxes.

The boxes are transferred f rom the RIHL to the RM0F and stored prior to shipment to an approved burial site.

1.3.1.2 Safety Considerations The probability of overexposure to direct radiation is very small, as the fuel assemblies are well shielded during transport and disassembly. and the workers are continually monitored with personal monitoring devices and film badges.

Since the dose received is the oroduct of the dose rate and the time of exposure, the individual's total dose can be regulated and kept within the specified limits by limiting his time spent in the radiation field.

Estimated exposure rates are not expected to exceed 2.5 R/h for the unshielded incoming fuel assembly and 1 R/h for the unshielded repackaged clad fuel.

The fuel s_ >

exposure rates are measured at 6 f t.

Accidents that could cause exposure to high radiation fields are of low probability due to engireeering safeguards and continuously monitoring radiation-level alarm systems.

I Transfer and storage of the bins will be in safe diameter containers, cont idering optimum moderation. These canisters exclude moderation.

l The water drip system, used in the assembly cutting process, contains several key features for criticality considerations.

First is the one-quart water supply container c:ed in conjunction with the saw.

The water is sup-plied to the saw in a slow-drip process, such that approximately 5 to 10 assemblies can be cut with one quart of water.

The dimensions of the collec-tion tray are such that a layer of fuel rods cannot collect in the tray with the water.

The water collected in the tray f rom the drip system is allowed to evaporate prior to having the saw chips vaCJumed out of the tray.

The wet vacuum is positioned a distance of at least 3 f t f rom the fuel and/or the col-lection tray.

If it is necessary to vacuum mixtures of water and chips, or Rl/R088-206 1-23

otherwise introduce water to the vacuum cleaner, the water level in the vacuum b]

cleaner is periodically inspected and, if the level reaches 4.5 liters, the vacuum is emptied and the contents solidified.

1.3.2 Tvoical Fuel Decladdina at the Hot Laboratory (8uildina 020) 1.3.2.1 Process Description Following is a process description of the SEFOR fuel decladding at the Rockwell International Hot Laboratories (RIHL), which is typical of the opera-tion currently planned for that facility.

The SEFOR fuel to be declad consists of 648 rods of the following types:

1.

626 standard rods contal' ing 18% Pu n

2.

12 instrumented fuel rods 3.

3 instrumented fuel rods with short fuel section 4,

7 rods containing 25% Pu.

.O The standard fuel is in the form of Pu0 -UO cylindrical pellets 2

2 (18.7% fissile plutonium 0.875 in, in diameter and ~0.64 in. long).

There are 55 pellets in each of the standard elements.

The fuel rods are clad with stainless steel and 49 5/8 in, long overall.

Only three rods, or the fuel from the three rods, containing ~600 g of Pu for standard rods (~805 g for the 25% Pu fuel rods), are in process with-in RIHL at any one time.

Storage of the remaining fuel inventory is at the Radioactive Materials Disposal Facility (RMDF), a DOE-owned facility operated by Rockwell International, inventory control of the SNM is carried out according to approved NMN procedures.

One on-site shipping canister is used to transport the fuel rods from the RMDF to RlHL.

After decladaing, the fuel pellets are returned to the RMDF in transfer cans.

Nonfissile material, scrap, and waste f rom the operation are returned to the RHOF from RIHL in appropriate container.

Rl/R088-206 1-24

r' x The decladding operation is completed within alpha boxes installed in one of the shielded RIHL rooms.

These boxes are maintained at a negative pressure x_-

with respect to the other areas by means of an independent pressure control system.

If necessary, the alpha box can be supplied with a continuous flow of nitrogen gas.

The decladding operation consists of a cutting operation at the end of the fuel rod and removal of the fuel pellets, other internal pieces, and cladding scrap to appropriate containers.

The fuel pellets f rom each fuel rod are placed into the transfer can and sealed, checked for contamination, weighed to overcheck contents prior to removal for the alpha box, and prepared for the return transfer to the RMDF.

Waste forms generated are fuel-rod hardware (cladding and internals),

plastic, Kim-towels, gloves, and tooling. Waste is stored until convenient to transport to the RMDF for disposal per DOE-approved procedures.

1.3.2.2 Safety Considerations h)/

Precautions are taken to limit exposures to personnel involved in hand-x, ling the fuel rods by follcwing standard health physics procedures.

A radia-tion safety representative monitors the operation to assure personnel saf ety and carries out the necessary surveys to assure the control of alpha contami-nation and radiation levels during and after each operation.

Criticality safety considerations show that three fuel rods are safe in any configurations with limited moderation.

Alcohol and/or other moderating liquids are controlled to less than 500 mi in the alpha box at any one time.

5123Y/ paw

\\m-Rl/R088-206 1-25

r 2.0 ENGINEERED PROVISIONS FOR ABNORMAL OPERATIONS m

The systems provided to detect accidental releases of radioactive mate-rials and provide alarm signals to assure corrective responses are (1) the radiation alarm system (RAS), which includes the accidental criticality alarms (ACA) and high-radiation alarms (HRA) as a subsystem to the RAS, and (2) high-airborne contamination alarms.

The systems provided to limit the release of radioactive materials are (1) the ventilation system, which includes the HEPA exhaust filters, (2) the enclosures (where required), such as fume hoods and gloveboxes, and (3) hot cells.

Additional features provided for abnormal operation are the fire detec-tion and suppression system, the emergency power supply, high-level glovebox features, and layout features for safeguards.

Since all operations with radioactive materials are manually controlled batch operaticns, there are no special requirements for safe shutdown of pro-cess systems, nor are any special systems needed to maintain a safe shutdown condition. Quantitles in all batch operations are limited to criticality-safe amounts or are in controlled-geometry containers.

Limited quantitles that could become involved in fires have been considered in the development of the accident consequences.

In Section 2.1, the standards, specifications, and safoguards features which apply to these systems are abstracted f rom the license technical speci-fication in the technical information supplied to support the relicense application.*

ESG-82-33, ' Health and Safety Sections for Renewal Application of the Special Nuclear Materials License SNM-21, Docket 70-25, Issued to Energy Systems Group of Rockwell International," Revised June 5, 1984 O

RI/R088-206 2-1

l 73 In Section 2.2, the adequar.y of the engineered provisions is discussed.

l U 2.1 STANDARDS, SF5CIFICATIONS, AND SAFEGUARDS FEATURES OF SYSTEMS TO ACCOMMODATE ABNORMAL OPERATIONS 2.1.1 System to Detect Accidental Releases and Provide Alarms l

l 2.1.1.1 Radiation Alarm Systems l

l An RAS is required in accordance with standards listed immediately below.

l The system must consist of radiation detectors installed at specified loca-tions to actuate an evacuation alarm at the affected locations nearby.

Detec-tor alarm components of an RAS installed to meet regulatory requirements for accidental criticality are called accidental criticality alarms (ACA).

Detec-tor alarm components installed to meet internal operational requirements for accidental high radiation are called high radiation alarms (HRA).

2.1.1.1.1 Standards for Accidental Criticality Alarms O

These standards are:

1.

An ACA is required as part of the RAS for each area in which 500 g or more of contained U-235, 300 g or more of contained

(

U-23') or plutonium, or any equivalent mixture is handled, used.

l or stored With the following exceptions:

a.

Where the total quantity of nuclear fuel in an area is no more than 45% of the minimum critical mass under condi-tions of full reflection and optimum water moderation for i

the enrichment and most restrictive chemical or physical form and configuration.

i l

b.

Where only a single f uel assembly may be present in an area.

2.

Each detector of an ACA system is considered to meet the cri-teria in (1) above if:

a.

Each operation, which it l '. Intended to cover, occurs within a combined shielding distance configuration equiva-lent to 120 ft in air, or O

Rl/R088-206 2-2

( 'g b.

Each operation, which it is intended to cover, occurs

')

within an enclosure (cell or vault) with suf ficient per-s manently installed shielding to ensure that personnel could not be exposed to a prompt radiation dose greater than 3 rem in the event of a 1018 fission energy re-lease.

In such cases, all of the following criteria must be met:

(1)

The prompt gama dose rate at the detector is 20 mrem /h or greater for an energy release of 1-MW-s (3.25 x 1016 fissions) over a minimum period of 0.1 s.

(2)

Personnel are not permitted within the enclosure un-less the access doot or aperture is open.

(3)

The dose rate at the detector with the access door or aperture open is 20-mrem /h or greater in the event that a condition of accidental criticality within the enclosure occurs which generates radiation levels of 300 rem /h at I f t from the source of the condition.

(4) With respect to adjacent areas, the enclosure is i

amintained at a negative pressure, while nuclear l

fuels exceding specified quantitles are present.

i

/T (5) A gas monitor with audible alarm is provioed in the (j

enclosure exhaust system, with suf ficient sensitiv-ity, and with an appropriate alarm set point to pro-vide a warning if the concentration of radioactive gas in the exhaust system exceeds 10-4 vC1/cm3 c.

The alarm set point is 20 mrem /h or less.

3.

The specifications of 10 CFR 70.24(a)(2) may be applied as the minimum criteria for an acceptable ACA.

This ser.tlon allows the use of a single detector to monitor a fuel handling area.

2.1.1.1.2 Standards for Hiah Radiation Alarms These standards are:

1.

An HRA is required as part of the RAS for each area in wh'.ch personnel may be accidentally exposed to a steady-state source from which the dose rate at i ft exceeds 300 rem /h or in which personnel may be accidentally exposed to a pulsing source f rom which the dose at 1 ft exceeds 150 rem / pulse.

If a determina-tion is made by R&NS that equivalent radiation protection is l

4 provided in another manner, the HRA requirement may be waived.

i O

i RI/R088-206 2-3 1

2.

At locations requiring both an ACA and an HRA, a single detec-gw) t tor must be used for both purposes.

The set point and distance criteria for ACAs are those established in the previous section.

3.

For areas in which only an HRA is required, any detector of the RAS may be used to satisfy this requirement, provided first that the provisions of (4), imediately following, are met and, second, that the detector automatically actuates the RAS alarm in the area requiring an HRA.

4.

Except as noted in (a), (b), and (c), imediately following, the set point for an HRA is maintained at a dose rate level not exceeding twice the peak normal working background, a.

If the peak normal working background is less than 20 mrem /h, the twice-background criterion is waived, b.

The upper limit for an HRA set point, as affected by dis-tance, is established by the following equation which is satisfied at each installation:

S = 5 x 1040-2F where S = alarm set point (rem /h) 0 = distance between the detector and the most likely location of the source under accident conditions (ft)

F = factor (less than unity) by which radiation f rom the accident source would be attenuated by material between the detector and the most likely location of the source.

The most likely location is to be determined by the affected operating department in conjunction with R&NS.

c.

At locations requiring both an ACA and an HRA, the set point is determined by the HRA standards or, as an upper limit, by regulatory standards.

5.

The location of each HRA is to be determined by both the af-fected operating department and R&NS.

The detector must be in-stalled in a location f avorable for the detection of radiation originating at the most likely location of an accident and in conformance with Standard (4)(a), imediately preceding.

O Rl/R088-206 2-4

('

2.1.1.1.3 RAS Performance Standards

()S The standards are:

1.

Each detector has a response time of 3 5 or less at 20 mrem /h.

2.

The alarm is a siren sound used exclusively to notify personnel to evacuate.

3.

The alarm is clearly audible in all affected areas.

4.

Permanently installed systems must provide notification of failure (power or sensor) to the Protective Services Control Center (PSCC), but such f ailure must, in no way, af fect the operation of other systems.

5.

Permanently installed systems are maintained and calibrated at 3-monthsintervals and repaired as necessary.

6.

Portable, temporarily installed systems are used whenever nuclear fuel is to be handled in quantities greater than those listed under Item 1 of 2.1.1.1.1 of this section in an area not within the coverage of an operable, permanently installed sys-tem.

Portable systems provide automatic capability of a siren

(]

alarm for evacuation.

'd 7.

Portable systems are used only unt'l (1) a permanent installa-tion can be completed when the handling operation is long term.

(2) complation of the operation when it is of short duration, or (3) the permanently installed system normally providing such coverage is repaired.

8.

The portable units provide notification of sensor failure (audible) either locally or to the PSCC.

9.

Batteries are changed at appropriate intervals to assure con-tinuous operation of the unit, and other maintenance, repair, and calibration must be accomplished at least once every 3 months for portable systems.

10.

Sensors are relocated only if all of the above standards are satisfied after relocation.

O RI/RD88-206 2-5

2.1.1.2 Standards for High-Airborne Contamination Alarms These standards are:

1.

Constant air monitors with audible alarms are required in areas where there is a potential for an accidental airborne radio-active material concentration of sufficient magnitude to cause an inhalation exposure which, in I hour, would exceed the ap-propriate occupational exposure standard established for 1 week, i.e., 40 RCG averaged over i hour.

2.

Airborne contamination alarms are clearly audible throughout the local area.

3.

Sensitivity and preset alarm points are such that the air mon-itor is capable of warning workers in a posted area in time to prevent inhalation exposures in excess of the appropriate time-weighted occupation exposure standards.

4.

To the extent practical, monitor inlets are located between the workers and the potential source of airborne contamination.

Alarming air monitors are kept in the following locations:

je(a,Geo 1.

RIHL Operatina Gallery--Sensitivef o both alpha and beta activ-t O

ity, with alarm points cor Eding to 3.9 x 10-10 vCi-h/

cm3 for alpha activity or 6.500JMPC-hours for plutonium re-leased to an unrestricted area and 1.5 x 10-9 vCi-h/cm3 for beta activity or 50 MPC-hours for Sr-90 released to an un-restricted area.

Dispersion to the nearest boundary would re-duce this exposure by a factor of 4,000, neglecting filtration prior to release.

Filtration would further reduce this by a factor of at least 2,000.

The resulting off-site exposure to alpha activity would be less than 0.0008 MPC-hours (Pu), and the exposure to beta activity would be less than 0.000006 MPC-hours (Sr-90) at the monitor alarm point.

2.

RIHL Service Gallery--Sensitive to beta activity, sith alarm point corresponding to 1.5 x 10-9 vCi-h/cm3 or 250 MPC-hours for Sr-90 released to an unrestricted area.

Dispersion to the nearest boundary would reduce this exposure by a factor of 4,000, neglecting filtration prior to release.

Filtration would further reduce this by a factor of at least 2,000, so that the maximum off-site exposure would be less than 0.00003 MPC-hours at the monitor alarm point.

3.

RIHL Basement--Sensitive to beta activity, with alarm point corresponding to 7.5 x 10-9 uCi-h/cm2 or 250 MPC-hours for Sr-90 released to an unrestricted area.

Dispersion to the O

RI/RD88-206 2-6

nearest boundary would reduce this exposure by a factor of

,o(')

4,000, neglecting filtration prior to release.

Filtration would further reduce this by a factor of at least 2,000, so that the maximum off-site exposure would be less than 0.00003 MPC-hours at the.mnitor alarm point.

4.

RlHL Exhaust Stack (Particulate Monitor)--Sensitive to beta activity, with alarm point corresponding to 1.5 x 10-9 pCi-h/cm3 for beta activity (50 MPC-hours for Sr-90 released to an unrestricted area).

Dispersion to the nearest boundary would reduce this exposure by a factor of 4,000, so that the beta exposure would be less than 0.01 MPC-hours at the monitor alarm point.

An alpha-sensitive monitor is provided on the portion of the exhaust system installed for high-alpha activity operations.

The alcrm point corresponds to 2.5 x 10-11 pCi-hours /cm3 (400 MPC-hours for Pu-239).

Dispersion to the nearest boundary would reduce this exposure by a factor of 4,000, so that the resulting alpha exposure would be less than 0.01 HPC-hours (Pu).

Also, a gas activity monitor is provided, when the possibility of a noble gas release exists, with alarm point corresponding to 1.9 x 10-5 vCi/cm3 or 65 MPC for Kr-85 released to an unrestricted area.

Dispersion to the nearest boundary would reduce this concentration by a factor of 4,000, so that the

T maximum concentration would be less than 0.02 MPC at the

)

monitor alarm point.

For comparison, the exposures (concentration at the exposed individual x time) corresponding to threshold Protection Action Guides (PAGs) and the off-site exposures corresponding to the alarm set points are:

1 Threshold l

PAG Alarm Set Points l

(NPC-Hours)

(MPC-hours)

-3 EU (3 rem to the lung) 11 x 10 1 x 10'

- 1 x 10

~4 Pa (3 rem to the bone) 6 x 10 1 x 10

- 1 x 10 l

Sr-90 (3 rem to the bone) 7 x 10 6 x 10

- 1 x 10'#

3

-6 Cs-137 () rem to the whole body) 50 x 10 4 x 10'I - 6 x 10 3

3

-6

-3 l

l-131 (5 rem to the thyroid) 28 x 10 2 x 10

- 3 x 10 Exposure to airborne radioactivity in an unrestricted area is authorized by 3

[

10 CFR 20.106 up to 8.7 x 10 MPC-hours.

1 l

Rl/R088-206 2-7 L

Additional data related to radiation exposure and airborne releases are provided by a high-pressure ion chamber (Reuter-Stokes RSS-111) installed to the east of the RIHL at Santa Susana.

Thermoluminescent dosimeters (TLDs) and air samplers are placed at several locations at Santa Susana, 2.1.2 System Provided to limit Release of Radioactive Materials 2.1.2.1 Soecification for Ventilation Systems Ventilation systems meeting Specifications 1 through 8 below are provided in posted areas, if necessary, under normal conditions to maintain radioactive material concentrations in air at sufficiently low levels to prevent inhala-tion exposures in excess of occupational standards established in 10 CFR 20*

or in CAC Title 17,* whichever is applicable. Ventilation systems, serving l

i areas housing high-level gloveboxes of this section, also meet Specifica-tions 9 through 11. Ventilation systems required for purposes of controlling potential accident consequences off site also meet Specifications 12 through I

16.

These specifications are:-

j 1.

The system provides a minimum of six air changes per hour in

[

the room (s) it serves.

2.

The air flow is directed from areas of lesser contamination potential toward areas of greater contamination potential.

3.

Air from posted areas is filtered as necessary and/or exhausted I

through a stack of sufficient height to assure that regulatory standards for radioactive material concentrations in air in unrestricted areas can be met under normal operating conditions without imposing otherwise unnecessary limitations on opera-t tions.

In-line filter bank etficiency testing is performed, using dioctyi sebacate (005) and a forward light-scattering photometer, as the necessity is indicated by stack sampling results.

10 CFR 20 is an abbreviation for Title 10, Code of Federal Regulations, Part 20; this regulation is appilcable to work with special nuclear materials.

CAC Title 17 is an abbreviation for California Administrative Code. Title 17; this regulation is applicable to work with radioactive materials other than O

special nuclear materials.

Rl/RD88-206 2-8

,S 4.

The ventilation system provides once-through air with no provi-i sion for recirculation.

V 5.

If continuous operations are conducted in an area requiring exhaust filtration, parallel duct and blower systems are pro-vided to facilitate maintenance, filter-changing operations, etc.

6.

Each blower system is equipped with an alarm device which auto-matically notifies the facility operators or the Protective Services Control Center in the event of blower failure.

7.

Water scrubbers, or an equivalent system, are provided in the exhaust f rom machine operations which generate sparks that could, ignite materials inadvertently accumulated in the venti-lation system or filters.

8.

Filter replacement is required when pressure gauges, as mont-tored b1 weekly, indicate 6 in, or more of water pressure dif-ferential across the filters.

Enclosure f ace velocities are periodically determined either directly or from the exhaust tystem volumetric flow rate.

9.

Where high-level gloveboxes are located, the ventilation system provides a minimum of 10 air changes per hour.

10.

Automatic valves and/or dampers are required in ventilation O

ducting as necessary 'to prevent backflow of plutonium and/or similar highly toxic radionuclides into unposted areas in the event of accidental pressurization of the ventilation system.

11.

Filters subject to contamination by plutonium and/or similar highly toxic radionuclides are self contained in a disposable housing or must be installed in housings provided with bagout capability.

12. Air f rom posted areas subject to accidents that could severely contaminate the air, such as accidental criticality, nuclear f uel, or radioactive v.aterial fires, is filtered as necessary and exhausted through a stack of suf ficient height to assure that radiological consequences of f site will not exceed those published by the NRC as guides for nuclear reactor siting.

Where filter systems are required for this purpose, the follow-ing specifications apply:

a.

Minimum Performance Specifications. Ef ficiency--The filter bank ef ficiency f or particles of 0.8-micron diameter is 99% for uranium and low-toxicity radionuclide areas using a standard ' cold" 00S test.

For plutonium and high-toxicity radionuclide areas, the filter media ef ficiency O

Rl/RD88-206 2 -9

is 99.95% for particles of 0.3-micron diameter, and the o) filter bank efficiency is 99.95% for particles of

(

0.8-micron diameter.

b.

Minimum Performance Specifications. Fire Resistance--All filters are constructed of fire-resistant materials.

For uranium and low-radiotoxicity radionuclide areas. the fil-ters are capable of continuous operation at 250*F with no loss in filtration efficiency.

For plutonium and high-radiotoxicity radionuclides, the filters are capable of operation for 5 min at 700 1 50'F with no loss in fil-tration efficiency.

Filters used specifically for mate-rial recovery do not need to satisfy these requirements if utilized in conjunction with filters for radiological con-trol purposes, c.

Minimum Installation Specification--The filters are lo-cated at a sufficient distance from the working areas to assure the maintenance of filter integrity under antici-pated accident conditions.

The filters are protected by prefilters or systems adequate to remove entrained incan-descent particles.

13. A stack monitor with audible alarm is provided if blower shut-down is required to limit off-site radiological consequences under accident conditions.

ph 14.

Redundant blowers are required for exhaust systems used to limit radiological consequences off site.

15.

Automatic emergency exhaust cooling systems are required for ventilation systems which are provided as necessary to prevent dange to filter media or to prevent bonding caused by high-temperature exhaust air following accidentc1 criticality or radioactive material or nuclear fuel fires, etc.

16.

The efficiency of facility exhaust filter bank systems to con-trol of f-site radiological consequences under emergency condi-tions is assessed by in-line testing, utilizing "cold" dioctyi sebacate (005) and a forward light scattering photometer at filter installation / replacement or, as a minimum, at a testing frequency of once per year.

l l

Rl/R088-206 i

2-10 l

i 2.1.2.2 Soecifications for Enclosures The specifications for enclosures are listed below:

1.

Enclosures, such as fume hoods and gloveboxes, are provided for all operations involving radioactive materials or nuclear f uels which, under normal working conditions, generate radioactive material concentrations in air in excess of occupational stan-dards.

Refer to Specification 14 for criteria governing the type of enclosure to be used for the more radiotoxic nuclides.

i 2.

The minimum average face velocity for a fume hood and for an opening in a special enclosure is 100 ft/ min.

For dusty opera-tions, the minimum is 150 ft/ min.

1 3.

For machining or other operations which conid impart high speed i

to a dust particle, the average face velocity at the enclosure opening is at least 150 ft/ min.

L 4

4.

Glovebores are required for work with any radioactive material l

l for which the quantity, specific activity, radiotoxicity, physical and chemical form, and existing or potential environ-

[

ment (thermal, pressure, chemical, etc) could, in combination, create potential hazards of such magnitude that fume hoods and i

special enclosures would offer insufficient personnel protec-O tion.

l I

5.

Separate minimum requirements are established for gloveboxes l

according to use. High-leve? gloveboxes are required for work l

with large quantitles of plutonium, other radionuclides of t

l I

similar radiotoxicity, and for less radiotoxic nuclides in such t

large quantitles that similar potential hazards ext' t.

Low-l s

j level gloveboxes are permitted for work with uranium and with sufficiently small quantitles of other radionuclides.

a I

6.

Gloveboxes are maintained at negative pressure with respect to the room housing them.

The minimum pressure differential i

requirements are -0.3 in, of water for high-level gloveboxes and -0.1 in, of water for low-level boxes. An exception to i

i this specification is that, if the use of the glovebox is for I

reasons other than control of airborne contamination, the nega-

[

tive pressure requirement need not be met.

i 7.

All glovebox exhausts are connected with exhaust systems meet-j ing all specifications listed above.

I 8.

Humidity-resistant, high-efficiency filters are also required at the inlet and outlet of high-level gloveboxes.

1 i

}

Rl/R088-20C 2-11 4

1 i

9.

An emergency exhaust capability, automatically actuated by loss m

(V of pressure differential, is provided for high-level gloveboxes i

to assure a minimum air flow rate of 150 ft/ min into a box through any two completely open glove ports.

This requirement may be waived if the exhaust system routinely provides a con-tinuous exhaust capability meeting the flow rate criterion.

10. A nanually actuated emergency exhaust capability is provided for low-level gloveboxes to assure a minimum air flow rate of 100 ft/ min into a box through one completely open glove port.

This requirement may be waived if the exhaust system routinely provides a continuous capability meeting the flow rate criterion.

11.

The maximum glovebox leak rates, with the boxes at -0.6 in, of water with respect to the room, are 1% of the glovebox volume in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

12. An inert atmosphere is required in a glovebox if necessary to prevent hazardous chemical reactions.

13. Glovebox shielding, either structurally mounted or internally arranged, is required if necessary to ensure that regulatory standards for external dose are not exceeded.

Remote systems are required if the hand dose is limiting and the radiation levels preclude glove work without exceeding regulatory ex-ternal dose standards for extremities.

W()

14. Radioactive materials for which the RCG in air is less than 4 x 10-10 uCi/cm3 are considered to be highly radiotoxic if the specific activity is sufficiently high that the RCG, ex-pressed in terms of mass, is less than 4 x 10-8 ug/cm3 For these radionuclides, tyoes of enclosures are specified according to the radioactive material quantitles involved and according to the operations to be performed.

These specifica-tions appear in Tables 2-1 and 2-2.

Table 2-1 provides basic quantity limits, while modification factors for these limits appear in Table 2-2.

The enclosure modification factors are contingent upon the type of operation involved.

To establish the applicable modification factor:

(1) the particular opera-tion (s) to be conducted and the quantity of material to be in-volved in the operation (s), are first established, and (2) the type of enclosure required is then determined by entering Table 2-1 with the product of the quantity of material to be involved and the modification factor.

The modification factor is selected f rom Table 2-2 on the basis of the type of opera-tion to be performed.

8 RI/RD88-206 2-12

Table 2-1.

Enclosure Requirements for Highly p

Radiotoxic Nuclides Q

Quantity Enclosure

>0.1 - 100 9C1 Fume Hood

>100 9C1 but $10 mci Low-level Glovebox

>10 mci High-Level Glovebox Table 2-2.

Modification Factors for Enclosure Requirements Operational Procedure Factor Normal Operations 1

Precipitation Filtration or centrifuging Solvent extraction in mixer settlers Chromatography Pipetting (not by mouth) or titrating active solutions Cleaning and Jegressing Storage (temporary) 0.01 Simple Wet Operations 0.1 O;

Diluting stock solutions for use b

Washing precipitates Complex Wet Operations with Risk of Spills 10 Distillation procedures Solvent extraction in pulsed column Sampling and transfer of solutions Evaporation to dryness using heat Simple Dry Operation 10 rusing procedures for preparation of solution Fluorination Transfer of dried precipitates Complex Ory and Dusty Operations 100 Machine or hand crushing Machining or sawing metals Steving Vigorous mixing by machine Melting Operations 1000 0

Rl/R088-206 2-13

15. All plutonium operations where the material is not encapsulated are performed in enclosures providing two stages of high-efficiency filtration. However, only the final stage of fil-tration will require testing upon initial installation and upon each replacement to assur6 conformance with the above require-ments.

2.1.2.3 Standards for Hot Cells Safeguard requirements for hot-cell operations are listed below:

1.

Hot cells are required for work with radioactive u terials if radiation levels and potential air contamination levels pre-clude work accomplishment in manipulator glove boxes without exceeding regulatory radiation protection standards.

2.

Cells are maintained at a negative pressure with respect to adjacent areas if potential airborne radioactive material har-ards exist. The normal pressure differential requirement is

-0.5 to -0.1 in, of water; under certain operating conditions, the differential pressure can be administratively controlled from -0.5 to -0.1 in, of water.

This condition exists when an attempt is being made to establish an inert atmosphere for operation with pyrophoric s terials, and it is necessary for O

operations personnel to feed nitrogen continuously.

These requirements my be waived during filter changing, ventilation system maintenance, and personnel entry into a cell.

3.

The ventilation system for a cell in which potential air con-tamination hazards exist may be shut down for filter changing and/or maintenance only if cell operations are discontinued and all cell openings are sealed.

4.

During personnel entry into a cell in which potential air con-tac.ination hazards exist, a linear air flow rate of 100 ft/ min is required across the door face and into the cell.

5.

Hot cells are to be used for work with materials, such as tr-radiated plutonium, which, in the absence of severe radiation level, would require enclosure to minimize the spread of con-tamination.

6.

Cell exhausts are connected with exhaust systems meeting all specifications listed in 2.1.2.1.

7.

An inert atmosphere is required for a cell if necessary to pre-vent hazardous chemical reactions.

O RI/RD88-206 2-14

l 8.

Shielding is required in ill directions to assure that (1) per-m

)

sonnel working full time in adjacent areas are not subjected to

.(k/

an annual whole-body dose greater than 5 rem; (2) the dose rate outside a facility or room in which the cell is housed is suf-ficiently low that personnel on the facility premises could not be subjected to a dose rate greater than 2 mres/h; and (3) per-sonnel offsite could not receive a dose greater than 2 mrem in I

1 h, or 100 mrem in 7 consecutive days, or 0 5 rem in 1 year.

9.

Electrical and mechanical interlocks and associated audit,le and visual alarms are required as necessary to ensure that person-nel are not exposed to inadequately shielded sources within a cell.

2.1.3 Additional Features Additional safeguard featVh are provided for protection in the event of abnormal operations.

The systems provided and safeguard feature are de-scribed below, i

2.1.3.1 Fire Detection and Suporession System These safeguards are listed below:

L 1.

Important structures are equipped with automatic suppression and/or detection systems based on zone occupancy evaluation, i

Where criticality safety for operations within a facility is based on limited water moderation or reflection, alternate sys-tems shall be provided (e.g., the production of combustion detection systems. inert gas system, special automatic pre-action system approved by the fire protection engineer combined with administrator controls).

2.

Suppression and/or detection systems are designed to sound a local audible alarm while simultaneously transmitting an alarm to the proprietary system console located in the Industrial Security Control Center.

3.

Areas are equipped with portabie ftre extinguishers and located i

in conformance with NFPA Standard 10, 4.

High-level gloveboxes are provided whth approved 1., sx thermal-detection devices with local audible alarm and simul-taneous alarm to the Industrial Security Control Center propri-etary system console.

l l

O Rl/R088-206 2-15

r 5.

Each high-level glovebox is equipped with an individual fire

(

extinguisher appropriate to the hazard and located adjacent to or within the glovebox, or dry-powder extinguishers with special piercing nozzles for penetrating gloves are located in the glovebox room.

6.

Suitable fire extinguishers are-located in the immediate vicin-ity of each low-level glovebox; procedural or design provisions are made for the use of these extinguishers without magnifying inhalation hazards, it 2.L,3.2 E7.ariency Power Suppiv Emergency power safeguards are listed below:

1.

A secondary, emergency power supply 4 prrvided for areas in which high-level gloveboxes are used c.ad for hot cells in which potentially severe airborne contamination hazards exist.

The units have a rating capaole of delivering full load at full voltage and frequency in approximately 30 seconds to the fol-lowing systems in the areas where they are required:

(1) the ventilatica system, (2) ac-powered radiological instrumenta-i tion, (3) instruraentation required for f acility status informa-fq tion, (4) all safeguards circuits, and (5) minimal facility

(/

illumination units.

2.

Emergency power units are required for other facilities only if necessary under eniergency conditions to limit of f-site rartio-logical consequences to within those published by the NRC as guidelines for nuclear reactor siting.

2.1.3.3 Hiah-l.evel Glovebox Desian Features These safeguards are listed below:

1.

Gloveboxes are constructed of 12-gauge (minimum) Type 304 stainless steel or equivalent with 3/8-in. plexiglass viewing I

windows or equivalent.

2.

Gloveboxes are constructed to withstand the static and live loads imposed during normal processing and rsanufacturing, with a factor-of-four safety margin.

3.

Glovebox floors are capable of sustaining the ef f ects of a nuclear fuel fire in addition to a reagent fire, without loss 1

of containment.

i l

Rl/RD8b-206 2-16 i

4.

Each glovebox is provided with instrumentation capable of acti-p

- (v) vating an alarm system in the event of fire, sufficient glove-box negative pressure, or excessive negative pressure.

5.

Gloveboxes are equipped with one or more access ports designed to permit the entry and removal of objects and mater'.als with-out directly exposing the glovebox atmosphere to the room atmo-sphere and without exposing contaminated surfaces to the room atmosphere.

2.1.3.4 Structural Intearity and Shieldina All facilities are constructed according to the Uniform Building Code as applied to industrial buildings.

This provides satisfactory protection against fire, earthquake, and structural collapse.

None of the operations involves any sigaificant amounts of flammable or explosive material.

Except for the hot cells, no engineered shielding is provided.

However, the building exterior and interior walls provide considerable attenuation of the prompt gama and neutron radiation associated with an accidental critical-ity.

The hot cells comprise a monolithic concrete structure that is quite O

resistant to any accident conditions.

The shielding effect of all the struc-tures is of significance only during the burst of radiation occurring within the first few seconds of a criticality event.

2.2 ADEQUACY OF ENGINEEREO PROVISIONS The enginesred provisions have provided and continue to provide the pro-tection for the ongoing licensed nuclear activities at Rockwell International.

During this period, no accidents have occurred that were not adequately con-trolled by the described engineered provisions.

Neither past accident analy-ses nor the current analyses reported in this plan have identified the need for either additional enginaered safeguard features or modifications to the existing features, d

5123Y/ paw O

Rl/RD88-206 2-17

l 3.0 CLASSES OF RADIOLOGICAL CONTINGENCIES V

In Section 3.1, the classfication system for accidents involving poten-tial or real radiological releases is described and defined.

In Section 3.2, guidance is developed for Rockwell employees to place accidents in the proper classification category based on known and readily observable facts with planned implementing responses.

In Section 3.3, the postulated accidents that can occur, with the results of the analysis of their consequences, are de-scribed.

They are then placed in the category in which they would fall if they should occur.

3.1 CLASSIFICATION SYSTEM This radiological contingency plan considers four subclasses of accidents that require offsite notification.

These categories are defined below and are categorized so as to assure the proper response to a possible or real radio-logical release.

It is recognized that any given event may be reclassified to a lower or higher subclass as inforr4.-

>n about the event is obtained or as the result of subsequent occurrences, 'ing an event.

The plans herein are a part of the total Rockwell plan mablished for handling the broader scope of industrial accidents that occur in any manufacturing, research, or development activity.

The subclasses of events are:

1.

Notification of Unusual Event II.

Alert III.

Site Area Emergency IV.

General imergency O

RI/R088-206 3-1

r These can be defined as follows:

N])

I.

Notification of Unusual Event--An event where there is a poten-tial degradation of the level of the safety of the plant with a significant potential for local off-site release of radioactive material.

II.

Alert--An event that involves an actual or potential substan-tial degradation of the level of the safety of the plant with any radiological releases limited to small portions of the EPA Protective Action Guideline exposure levels.

III. Site Area Emeraency--An event that has occurred involving actual major f ailures or occurrences and of f-site releases are not expected above the EPA Protective Action Guidelines ex-posure levels except near the site boundary.

IV.

General Emeraency--An event that has occurred that has resulted in actual or imminent releases that can be expected to exceed the EPA Protective Action Guidelines exposure levels of fsite for more than the insnediate site area.

3.2 EVENT CLASSIFICATION AND RESPONSE V

The first categorization of an incident or event will be by those employea who are in the immediate vicinity and intimately involved.

During of f-hours, this will be the Industrial Security and Protective Services per-sonnel.

During on-duty hours, the personnel assigned at the f acility involved will make the judgment based on the known activities at that facility end the readily observable facts.

This approach is predicated on the assumption that the kind of accident that has occurred may not be readily discernible but that t

the results of any event are observable, and the event classification and sub-l sequent response must be based initially on these facts.

In either case, the Protective Services organization is charged with the responsibility for notification and call-in of the Emergency Response Team and other emargency action personnel in accordance with implementing procedures, f

These implementing procedures are described in Appendix A.

l i

O Rl/R088-206 3-2

It may be necessary to reclassify an event as more information is ob-

'v tained and in the course of time as the event proceeds under control er other complicating factors become apparent to require escalation.

Table 3-1 has been developed for use by the Industrial Security and Pro-tective Services personnel.

It should be noted that these personnel will be required to call in the Emergency Response Team to assume responsibility for event classification and response, to take any life-saving actions that may be required, and to take those actions necessary to contain the event and main-tain the "status quo."

Table 3-2 has been developed for use by the involved facility personnel during on-duty hours.

Table 3-3 sumarizes the actions to be taken by the Emergency Response Team after the determination of the event classification by the employees in the imediate vicinity.

3.3 RANGE OF POSTULATED ACCIDENTS A comprehensive range of accident has been postulated for the licensed facility at SSFL.

These accidents include fire, eplosion, and accidental cr.icality without consideration for the credibility or likelihood of initi-ating events.

Rep.esentative accidents (as presented in "Emergency Plan for Atomics International Facilities Licensed Under Special Nuclear Materials License SNH-21," Al-78-14) are summarized in Table 3-4, showing the radiological con-sequences at the nearest site boundary.

In all cases, the nearest residence or substantially occupied 1ccation is at a considerably greater distance.

It should be noted that none of these accidents results in determination of a "General Emergency."

O RI/RD88-206 3-3 L.

O O

O i

l l

1 Table 3-1.

Decision Matri :-Radiological Contingency-Of f-Duty Hours i

Effectivity Santa Susan.a Building 020 (Hot Laboratory)

Field Labor-l atory Site i

Off-Normal Conditions Observed Action Take all necessary life-saving actions Fire or Energy Release (including near brush fire)

Take all necessary action to maintain m

g status quo u a 1$

Call in Emergency Response Team

+

Take all necessary life-saving actions Radiation or Criticality Alarms Isolate facility Call in Emergency Response Team Take all necessary life-saving actions Earthquake or Other Natural Disaster--With Suspected Structural Damage Isolate fccilities Call in Emergency Response Team 5125Y/jlm

n

^ :

O U-Table 3-2.

Classification Decision Retrix - Radiological Incidents - On-Outy Hours Effectivity: Building 020 RI Hot Laboratory-SSFL Energy Release nadiation Release Structural External Engihaered Requires or Equipment Safeguards Event Local Evacuation Local Damage Internal On-Site Off-Site Involved I.

Notification Yes tio Yes Yes(I)

Iso Ito of Unusua)

Yes No Yes Yes(1)

No IIe 5

Event Yes bio Yes Yes(I) 10 0 Yes

)

Yes 80 0 Yes Yes(l)

Ilo Yes YE

7 II. Alert Yes No Yes Yes(2) go, yes Yes leo Yes Yes(2)

Iso Yes Yes Yes Yes Yes(2) sto Yes (3) -

Yes Yes(2)

No yes III. Site Area Yes Yes Yes Yes(2) yes(1) yes Emergency Yes Yes Yes Yes(2) yes(1) yes (3)-

Yes Yes(2)

Yes(1)

Yes IV. General Yes Yes Yes Yes(2)

Yes(1) yes Emergency Yes Yes Yes Yes(2) yes(2)

Yes Yes Yes(2) yes(2) yes (2)-- g --

(1) Small fractica of EPA PAG exposure limit (2) Exceeds EPA PAG exposure limit (3) As a result of unspecified natural disaster 5125Y/jle

\\

/

s.

i l

1 1

I i

Table 3-3.

Action Guide to Radiological Events I

I Event Ismediate stotification

]

Classification Response Level Level Recovery and Restart Approval i

I.

Biotification of Emergency Response Tem Emergency Response Team Leader unusual Event l

Leader Director, NS&L Director,elS&L HS&E Representative Director, Indus-Director responsible for x

j q

trial Security facility FSIE Representative Director, F&lE m

3 l

4g Industrial Security NRC Region V Facility / Functional s

"o Management

Pubile Relations j

II. Alert As above As above plus As above j

top management III. Site Area As above As above plus As above plus top management Emergency state and/or local authorities i

leRC Headquarters IV. General Emergency As above As above As above plus IIRC requirements

$125Y/ paw l

4

Table 3-4.

Exposures Resulting from Airborne Releases (m) at Nearest Boundar!< (rem)

Whole Other Conti.igency Accident Body Thyruid Organs Class Fire in RIHL, Ar-85 6.5 x 10-6 nil nil Unusual released Fire in RIH., I-131 nil 0.31 nil Alert released Accidental criticality in RIHL 0.21 0.13 0.22 Alert (GI tract)

Accidental criticality in RIHL 0.21 0.44 0.22 Alert (irradiated fuel)

(GI tract)

The accident classed as "Unusual" would result in undetectable concen-trations of radioactive material (airborne or ground) of f-site, and assessment measurements would not need to be extended away f rom the innediate neighbor-hood of the affected facility. Accidents classed as "Alert" would be identi-fied by the detection of significant concentrations of airborne and ground-O,s deposited radioactivity, corroborated by knowledge of operations and accident r.onditions in the affected facility.

The following conditions were assumed in developing the consequences of the Postulated accident:

1.

Criticality:

a.

30 MW-s energy release 2.

Fission product und ftel release to air:

a.

Noble gases--100%

b.

Halogens--25%

c.

Particulates--l%

d.

U metal--0.24%

~0 e.

U oxide--6 x 10 %

.C f.

Pu metal--S x 10 '%

i

-6 g.

Pu oxide--6 x 10 %

lO RI/RD88-206 3-7 l

3.

Particulate filtration ef ficiency, 99%, with no decay in tran-(n) sit, no plateout w.J 4.

Fission product and fuel discharge to atmosphere:

a.

Noble gases--100%

b.

Halogens--25%

c.

Particulates--0.01%

-3 d.

U metal--2.4 x 10 g

~

e.

U oxide--6 x 10 %

f.

Pu metal--5 x 10-7%

g.

Pu oxide--6 x 10 %

5.

Shielding (attenu; tion of gansna and neutron radiation):

RIHL (Santa Susana)--42-in. high-density concrete 6.

Attenuation by air and heavy equiprnnt was neglected.

7.

Iodine inventory, in irradiated fuel at io. L only, 9.3 C1 1-131.

8.

Atmospheric dispersion was calculated for least diluting condi-tions at each distance, with wind speed assumed to be 1 m/s.

9.

Distance to nearest offsite point of exposure:

RIHL--381 m 10.

Stack discharge height:

RIHL--22.2 m

11. Oc:upancy time for exposure to direct radiation was assumed to be infinite at the Santa Susana boundary (a hypothetical resi-der ce).

For exposure to airborne radioactivity, occupancy time was assumed to be throughout the complete passage of the plume.

Many of these assumed conditions are, in fact, quite conservative and lead to a considerable overestimate of the exposures resulting from the hypo-thetical incident.

For example, the postulated release fractions are high estimates for a criticality accident involving a fuel solution and greatly overestimate releases from a metal system.

Decay in transit and plateout would reduce the amount of noble gases and halogens penetrating the filter.

Filter ef ficiencies are generally a f actor of 10 better than that assumed h e '. e.

At the distances involved in this study, attenuation of the neutron and garrina radiatim by air becomes significant.

The choice of 1 m/s wind speed is also conservative at several distances since the stability conditions required for maximum doses at these distances do not exist with such a low wind speed.

! O Rl/R088-206 3-8 l

i

/

Instruments are available throughout the site that permit measurement and evaluation of releases.

These instruments are accessible to members of Radi-ation and Nuclear Safety, who are trained in and familiar with their use.

The following instrumentation capability is available for prompt detec-tion and continued assessment:

1.

Radiation exposure rate:

a.

Radiation alarm system (RAS) detectors installed as part of the criticality alarm system, providing remote readout relative to whether the local exposure rate exceeds 20 mrem /h or not.

Can be read locally over range of 0.1 to 1,000 mrem /h.

Four units are associated with the RIHL.

b.

Reuter-Stokes high-pressure ion chamber RSS-111 with back-up battery power.

Provides current exposure rate, from natural background to 5 mrem /hr, record of past exposure rate, and integral exposure since last reset.

Normally read and reset at end of each calendar quarter.

One unit is at Santa Susana to the east of the RIHL.

p c.

Ludlum Na! scintillator rate meter, Model 12S Micro-R Q

Meter.

Provides measure of current exposure rate, f rom natural background to 3 mrem /hr.

Six units are assigned to Radiation Safety staff members at Santa Susana, d.

Thermoluminescent dosimeters (TLDs), Victoreen.

Provides integrated exposure since last readout in the range of 5 mrem to 1,000 rem.

Normally read out at end of each calendar quarter.

Six sets are located at Santa Susana.

2.

Airborne radioactivity:

a.

Installed Continuous Air Monitors--Eberline Alpha-3 air monitors are located in the RIHL when plutonium is being processed.

Eberline air monitors AMS-2 and Nuclear Measurements Corp.

AM-33P are located in the RIHL.

b.

Fixed Air Samplers--Fixed air sampler heads are located near most work stations processing unencapsulated radio-active materials in the RIHL.

The exhaust stacks of these facilities are also provided with fixed air samplers.

O Rl/R088-206 3-9

c.

Portable Air Samolers--Staplex high-volume (ac-powered)

[Q

\\

air samplers are available, as are Hi-Q Filter Products 12 Y de-powered high-volume air samplers.

3.

Suriace Contamination:

a, 1here are 46 ac-powered count rate meters available, with alpha scintillator detectors or pancake G-M probes.

There are an additional 42 portable battery-powered count rate meters with alpha scintillator detectors or pancake G-M probes.

b.

Surface snears can be counted for alpha and beta activity simultaneously by use of Nuclear Measurements Corp. ACS-7 sample counters (three at Santa Susana) or Canberra Low-Background counting systems.

4.

Radionuclide Identification:

a.

Gama spectrometry can be performed by use of high-purity germanium detectors with multichannel analyzers.

The manpower required for assessment and suppression is inanediately available or on call and is satisfied by current staf fing associated with the operations conducted under this license.

A major portion of the Operations, Protective Services, and Safety staf f reside within 1/2-h travel time from the site.

O Rl/R088-206 3-10

4.0 ORGANIZATION FOR CONTROL OF RADIOLOGICAL CONTINGENCIES

!d 4.1 NORMAL PLANT ORGANIZATION ine organizational structure important to the response actions during on-duty hours for the Rockwell International Hot Laboratory involved with NRC-licensed activities is described below.

As stated in Section 3.2, the first categorization of an incident or event will be by those employees who pro-cedurally have that responsibility and authority, i.e., the managers of the facility or their delegate.

Their action is to make this classification and inform Protective Services to initiate notification of the Emergency Response Team, who then assume further responsibility for response as described in Sec-tion 4.2.

Building 020 Rockwell International Hot Laboratory Assistant Manager and Nuclear Operations Manager, Report to:

Development and Test Manager Chief Engineer, Atomics International Division Director, Atomics International Ouring of f-duty hours, the responsibility for the initial response, as indicated in Table 3-1, rests with the on-duty site Protective Services field supervisor at the Santa SJsana site.

The site field supervisor reports to:

Director of Industrial Security Vice President, Human Resources and Communications The Rocketdyne Division organization chart in Figure 4-1 shows the rela-tionship of the Division Director, Atomics International, and the Vice Presi-dent Human Resources and Communications, to the President of the Division and other organizational entitles.

RI/R088-206 4-1

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Rl/R088-206 4-2

4.2 ON-SITE RADIOLOGICAL CONTINGENCY RESPONSE ORGANIZATION g

U The responsibility for the response to the radiological events rests with the Emergency Response Team leader.

During on-duty hours, the facility man-ager or his delegate is responsible for activating the Emergency Response Team by notifying the Industrial Security Control Center.

Prior to the arrival of the Emergency Response Teain, the facility managemeiit has the responsibility to contain the event, make the initial determination of the event classification, and notify the Industrial Security Control Center.

The security control cen-te.- initiates response by emergency action functions (reference Ap;:endix A).

These functions take the actions necessary to control the situation and pre-vent escalation.

During of f-duty hours, Protective Services will call in the Emergency Response Team as indicated in Table 3-1.

The Emergency Response Team leader will make the determination of the event classification.

In the interim period, the Protective Services personnel will take those actions necessary for lifesaving and to maintain and control the situation to prevent' escalation of the event.

The Emergency Response Team has been organized to be responsible for the response action for any of the four classes of radiological events.

Their responsibility begins on the arrival of the team leader at the location of the event and continues yntil the team agrees that the event is stabilized and no longer transitory and recovery can be planned and carried out in a logical straightforward manner.

At this time, the recovery ef forts are transferred f rom the Emergency Response Team to the f acility/ operation management.

The team leader and alternate have been delegated this responsibility by Rocket-dyne management, lhe team members and alternates represent Mealth, Safety & Environment; Industrial Security; Facilities and Industrial Engineering; and comunica-tions.

Members of the team and alternates are designated by the team leader O

RI/R088-206 4-3

with the approval of the functional directors involved and Rocketdyne manage-ment.

The team leader may also represent one of the functions required for the team membership.

The functional management for the facility where the event has occurred is also represented in the Emergency Response Team.

The managers of the Hot Cell have designated the team members and alternates who would be designated for an emergency event.

These members are also approved by Rocketdyne management.

4.2.1 Direction and Coordination The Emergency Response Team Leader (or alternate) will report to the Division Director, Atomics International.

The team members report to the tecm leader until the event is terminated by formal release of the involved facil-ity by approval to restart.

The duties of the team leader are to take charge of the emergency re-sponses, take necessary actions to terminate the event, minimize the event consequences, recover from the event, and take the necessary steps to stabil-ize the situation in preparation for the recovery planning.

The turnover to f acility/ operations management will be made when this has occurred.

This decision will be made on a case-by-case basis but will never be prior to ter-mination of the event.

The other team members will support the team leader and will have full recourse to the functional capabilities of the organization they represent on the team.

4.2.2 Emergency ResDonse Team Members Contingency Assignments The makeup of the Emergenc / Response Team is described above.

To reiter-ate, the team members are responsible to the team leader since they represent functional organizations with specialized capabilities the team can draw upon.

These capabilities are detailed below for each of the team members.

O Rl/R088-206 4-4

fm 4.2.2.1 Health. Safety & Environment (HS&E)Regresentative k

This team member and the organization that he represents provide the necessary radiological surveys and assessments, recor.iendations for contamina-tion control, work control procedures in contaminated areas such as protective clothing, air monitoring, dosimeter needs, and decontamination procedures and direction.

4.2.2.2 Facilities and Industrial Engineerina (F&IE) Representative This team member and the organization that he represents provide the necessary engineering assistance for repair and damage control and provide any temporary structure that may be needed for such needs as decontamination or contamination control.

Also, he would provide any necessary engineering assistance.

He would also provide post-event assessment and recovery assistance.

4.2.2.3 Industrial Security Representative This team member and the organization he represents provide fire protec-tion engineering / fire control, rescue operation, first aid needs, communica-tion needs, security and access control needs, and transportation (including ambulance service).

4.2.2.4 Facility Management Representative This team member is assigned from the organization responsible for the r

normal operation of the involved facility.

As such, he will provide up-to-date information on the status of the facility and of the operations underway im-mediately prior to the event.

He will also provide the personnel roster at the f acility, assure personnel accountability, and provide detailed f acility inf ormati,a needed for rescue operations.

He will provide necessary opera-tional and f acility information needed to contain the accident and for subsa-quent repair, damage control, and recovery.

O RI/R088-206 4-5

4.2.2.5 Communications Representative g) r'V This team member is assigned to the team leader to provide comunication capability to management and to the media according to established Rockwell policy.

4.3 0FF-SITE ASSISTANCE TO FACILITY Rocketdyne has written agreements and arrangements with private and public agencies to provide assistance in case of radiological emergencies.

These are summarized below.

4.3.1 Medical Treatment Facilities The three medical stations at Rocketdyne are staffed 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per day, Monday through Friday, by an occupational physician, certified physician's assistant (PA-C), five registered nurses, a certified X-ray technologist, and a licensed vocational nurse.

The medical doctor, the PA-C and the supervising nurse are available during normal working hours and on call at all times.

Rocketdyne has an association with Humana Hospital West Hills wherein the hospital has demonstrated and continues to demonstrate an appropriately trained professional staf f, up-to-date equipment, and facilities for the treatment of radiologically contaminated patients.

Both the Medical Director and the 01 rector of Nurses of the emergency room are REAC/TS trained (Radia-tion Emergency Action Center / Training Site) by request of the Rocketdyne Medical Director.

l The hospital is not only prepared to accept contaminated patients, but has participated in extensive readiness training drills with Rocketdyne in-volving hospital personnel and equipment.

O RI/R088-206 4-6

)

4.3.2 Medical Evacuation Service Rocketdyne has arrangements for emergency medical evacuation with Los Angeles Fire Department (LAFD) paramedic ambulances, two commercial ambulance services, and University of California, Los Aligeles (UCLA) Med-Star helicopter ambulance to transport radiologically contaminated patients to appropriate hospitals including Northridge Medical Center, Humana West Hills, UCLA Medical Center, and Sherman Oaks Burn Center.

4.3.3 Firefightina Backup The first response to fire alarms is by the Rocketdyne Fire Department and equipment located at the Santa Susana Field Laboratory.

The Ventura County Fire Department would provide the necessary firefighting backup.

The Ventura County Fire Department has participated in the SSFL emergency drills j

and is periodically briefed on the activities underway at SSFL in order to be l

better prepared to provide assistance when required.

l 4.3.4 Police Assistance The Ventura County Sherif f's Department has participated in emergency drills and is periodically briefed on the activities underway so as to be better prepared to provide police assistance.

Both the los Angeles Police Department (LAPD) and the Ventura County Sherif f's Department have included possible Rocketdyne requirements in their contingency planning documents.

The location of the Sheriff's Department is indicated in Figure 1-2.

The LAPD is concerned since the principal egress f rom the SSFL is directly into the City of Los Angeles.

O Rl/RD88-206 4-7 1

,ew 4.4 COORDINATION WITH PARTICIPATING GOVERNMENT AGENCIES b

The principal participating government agencies and organizations with action responsibilities for radiological emergencies are included in the Federal Radiological Emergency Response Plan (FRERP), which provides for Federal agencies to discharge their responsibilities during a wide range of peacetime radiological emergencies.

These agencies include the following:

a.

Federal Emergency Management Agency (FEMA) b.

Department of Agriculture c.

Department of Commerce d.

Department of Defense e.

Department of Energy f.

Department of Health and Human Services g.

Department of Labor h.

Environmental Protection Agency (EPA) 1.

Interstate Commerce Commission j.

Nuclear Regulatory Commission k.

Joint Nuclear Accident Coordination Center.

The responsibilities and capabilities of those entities are summarized below.

4.4.1 Federal Radiological Emergency Response Plan The Federal Radiological Emergency Response Plan (FRERP) is an agreement entered into by federal agencies of the United States Government, as lhted above.

The main purpose of the FRERP is to establish an organization and operating agreement to be used in the event of a rnajor accidental release or loss of control of radioactive material seriously endangering the public health and safety.

Through the FRERP, it is expected that econcalc and ef fec-tive use of federal agency resources can be made to mitigate the accidental radiation exposure of the public and individuals, minimize the spread of O

Rl/RD88-206 4-8

radioactive materials into the environment, and carry out countermeasures appropriate to the control and removal of radiological hazards.

The FRERP is intended primarily for handling radiological hazards that result from peacetime use of nuclear energy.

The plan provides the Federal Government's concept of operations based on specific authorities for respond-ing to radiological emergencies, outlines Federal policies and planning assumptions that underlie this concept of operations and on which Federal agency response plans are based, and specifies authorities and responsibil-ities of each Federal agency that may have a significant role in such emergen-cies.

The plan includes the Federal Radiological Monitoring and Assessment Plan. (FRMAP) for use by Federal agencies with radiological monitoring and assessment capabilities.

The Department of Energy is assigned the responsi-bility for developing FRMAP, which provides for a coordinated Federal response when emergency radiological assistance is requasted.

Although the FRMAP is a Federal agency plan, its objectives include the integration of Federal, State, and local government emergency capabilities.

Any incident involving radioactive material or a radiation hazard may be reported through many separate channels.

If an incident report or request for j

assistance is received by a local police or fire department, the department takes such action as is within its jurisdiction and abilities and requests t

state and/or federal radiological assistance as appropriate.

If the report or request is made to a state agency, the agenc/ contacts the appropriate local and 00E agencies for assistance, if needed. A request for radiological assis-tance may come dii $ctly to 00E f rom the general public.

00E makes the deter-mination whether the request should be referred to a state agency having jurisdiction and/or requests auxiliary assistance f rom appropriate federal agencies under FRMAP.

In any case, the DOE assures a response to emergency assistance requests.

O Rl/R088-206 4-9

L 4.4.1.1 Procedures for FRMAP Region 7 4.4.1.1.1 Responsibility for Coordination The San Francisco Operations Of fice of the DOE (00E-SAN) is assigned the authority and responsibility for developing the FRMAP for Region 7, which includes the States of California, Nevada, and llawall, and the Pacific Trust Territories.

DOE-SAN functions as the coordinator and liaison for the inte-grated off-site monitoring and assessment activities of the 000 and other Fed-eral agencies under the terms of FRMAP in response to incidents occurring within Region 7.

4.4.1.1.2 Responsibilities for Other Federal Agencies Specific authorities and responsibilities assigned to the participating agencies are listed below:

1.

Participating Federal agencies will develop plans and supporting O,s procedures at the regional level to integrate with FRMAP-7. Those plans will be consistent with any planning requirements placed on state and local governments and specific nuclear facilities or oper-ations for response to radiological incidents.

2.

Participating Federal agencies maintain facilities, equipment, and personnel to carry out their statutory responsibilities.

The use of these resources for radiological monitoring and assessment during an emergency will be coordinated under FRMAP-7 during response to rad-iological incidents.

3.

Federal agencies will make resources available under FRMAP-7 upon request to the extent that the agencies can continue to carry out their essential missions and statutory responsibilities.

4.

An agency that makes its resources available, although under the direction of 00E-SAN (or the EPA), does not place itself under the authority of the coordinating agency.

5.

Federal radiological response will be in support of and coordinated with that of state and local governments and owner / operators of f ac-tiltles and materials.

I i

O R1/R088-206 I

i

' 6.

00E-SAN will initiate FRMAP-7 following a request from state or local governments, another Federal agency, or when 00E-SAN, in coor-dination with 00E Headquarters, decides to respond under its statu-tory authority.

7.

00E-SAN will assist other Federal agencies and state and local gov-ernments with planning and training designated to improve local response capabilities and will support drills, tests, and exercises.

8.

Emergency actions may be taken by participating Federal agencies on their own authority in order to save lives, minimize immediate haz-ards, or gather information about the incident that may be lost by delay.

9.

Funding for each agency's activities in support of FRMAP-7 is the responsibility of that agency.

4.4.1.2 Participatina Aoency Functions and Capabilities The San Francisco Operations Of fice of 00E is the coordinating of fice for FRMAP Region 7.

Five radiological assistance teams in the San Francisco Bay l

area, t.os Angeles area, San Diego area, and Las Vegas area are available and can provide radiological, medical, and decontamination advice and services; they can also arrange aerial radiological surveys and meteorological services.

The Department of Agriculture, through the Animal and Plant Health In-spection Service, can provide limited capability in radiation monitoring and assist in removal from the market of radioactively contaminated meat and poultry.

The Department of Comerce, through the National Weather Service, pro-vides weather information and forecasts for use in planning protective action for relief programs.

Up-to-date weather reports and forecasts and special weather advice are available f rom the Weather Service Forecast Of fice.

O Rl/R088-206 4-11

The Department of Defense (000) is primarily concerned with radiological t

incidents in connection with 000 activities, but it has highly qualified Nuclear Accident Incident Control Teams which can respond to non-000 radio-logical incidents if requested by the FRMAP coordinator or duly constituted local authority on an as-available basis.

The Federal Emergency Management Agency provides general assistance, coordination planning, and support capability to the FRMAP programs and teams.

The Environmental Protection Agency has radiological capabilities which include portable monitoring and transportation, medical advice, laboratory facilities, whole-body counter, and highly qualified health physicists.

The Food and Drug Administration can provide assistance in radiation monitoring of foods, drugs, cosmetics, household substances, and therapeutic devices.

Laboratory facilities are also available for nonradioactive analyses.

The Interstate Comerce Commission can furnish technical knowledge rela-tive to land transportation of radioactive materials and can assist in locat-ing surface carriers to expedite emergency transportation of people and supplies to and from accident locations.

The Joint Nuclear Accident Coordination Center coordinates 000 and 00E response. While the center does not have radiological capabilities, it can request assistance f rom radiological assistance units in the 00E and 000 under the Joint 000 and 00E agreement.

Assistance may include transportation of response teams by the Air Force.

The Nuclear Regulatory Comission has substantial radiological monitoring Capabilities together with a number of managerial, technical, and profes-sional personnel who provide a significant manpower resource at the Region V office for FRMAP support.

O Rl/R088-206 4-12 l

~}

A complete description of the agency functions and capabilities relative N'

to the Assessment Plan is contained in the DOE Document, "Region 7 Federal Radiological Monitoring and Assessment Plan," February 1986.

4.4.2 00E Emeraency Radiological Assistance Team As indicated above, the SAN office of DOE has five radiological assis-tance teams available.

The membership of the team from the los Angeles area includes health physics and medical staff from Rocketdyne who have available the instrumentation, laboratory, professional staff, and associated facility capability for emergency response.

In addition, the other four teams are j

available as needed.

4.4.3 California State Radiologic K3alth Section The California Radiologic Health Section, located in Sacramento, is the agency responsible for the control of state-licensed radioactive material, and, in the event of an incident, the "California Radiological Emergency As-O sistance Plan' is implemented by the section and the participating agencies.

4 The section and the participating agencies have complete radiological monitor-ing capability and respond to requests for assistance for the 00E and the FRMAP.

4.4.4 Los Angeles County Department of Health Services I

The Los Angeles Division of Radiological Health in the Department of Health Services, located in Los Angeles, provides radiological monitoring and l

assessment capability on requests from the State Radiologic Health Section, 00E, and the FRMAP.

t l

5123Y/j lm O

R!/R088-206 L

4-13 i

i

n 5.0 RADIOLOGICAL CONTINGENCY MEASURES U

The notification of the Emergency Response Team for emergencies of all classes, for emergencies that occur either during operational or nonopera-tional hours, is described in Section 4.0.

The personnel required to support the Emergency Response Team for radiological controls and contingencios are notified and assembled as required by the Emergency Response Team.

5.1 ACTIVATION OF RADIOLOGICAL CONTINGENCY RESPONSE ORGANIZATION The Emergency Response Team will call in those personnel that have been designated to provide support radiological assessment capability.

These notifications to respond are made by in-plant telephone and PA speakers, per-sonal radio communicators, and commercial telephone, depending upon the time of day and the day of week.

Initial notifications are made to responsible facility personnel and radiation safety staf f, as identified on the emergency call lists, and appropt tate security and fire protection personnel.

As the extent of the incident becomes better defined, additional personnel are noti-fied as required.

No message authentication scheme is in use and none is planned.

5.2 ASSESSMENT

ACTIONS The personnel so ideratified to provide assessment capabilities to the Emergency Response Team can make determinations of the severity of any fire, seismic, or explosion-related damage, and the condition of emergency backup and engineered safeguards systems.

Also, they can provide the determination of the radiological conditions associated with the incident.

Personnel with radiation detection equipment survey the surroundings and, if appropriate, reenter the involved facility.

This reentry is done using personnel protective equipment, including self-contained breathing apparatus (SCBA) and protective clothing, to minimize the O

RI/R088-206 5-1

p potential hazards.

Backup teams are provided and communication with the

'd Industrial Security Control Centers is established.

For conditions indicative of a more severe event, additional personnel can be deployed and further-ranging measurements of radiation, contamination, and airborne radioactivity could be made.

These personnel consist of senior members of the Health, Safety & Envi-ronment Department.

Response and assessment involves retrieval of criticality dosimeters and TL0s in the event of an accidental criticality, reading the in-plant and on-site radiation monitors, recovering the fixed air sampler fil-ters, and performing the appropriate radiometric analyses.

Meteorological data are reviewed to determine the direction of any airborne releases and to estimate the concentrations and exposures that might occur under actual wea-ther conditions.

Wind speed and direction are obtained f rom sensors within 200 m of potential release points.

The stability class is determined by observing existing conditions.

The method and parameters that are used are described fully in "Meteorology and Atomic Energy."

For example, the disper-sion curves calculated for the accident analyses prosented in this plan are shown in Figure 5-1.

The relative concentrations are conservatively greater than those presented in U.S. NRC Regulatory Guide 3.34.

If appropriate, teams are sent out to determine surface deposition.

5.3 CORRECTIVE ACTION 3 Corrective actions by the Emergency Response Team are based upon initial and continuing assessment of the incident conditions by its support person-nel.

Health Physics personnel supply information on radiological conditions, i.e., answering such questions as whether or not there has been a release or spread of radioactivity.

If there has been a release, what are the levels of radioactivity where personnel are present to take corrective steps to mitigaie the consequences of the accident?

How long may these personnel stay? What are the levels of radioactivity at the facility and site boundaries?

Fire engineer and fire protection personnel supply information on structural damage O

RI/R088-206 5-2

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Atmospheric Dispersion for Limiting Working Inventories Rl/R088-206 5-3

and, when thw e are fires, give prognoses on containment.

Protective Services personnel give information regarding crowd control and personnel accountabil-x 1

ity (in cooperation with the involved facility's.3anagement), and relay infor-mation f rom thv communication network.

Informativn on the accident, such as what happened, the physical condition of the plant and processes as they know it, and potential dangers or circumstances that might lead to an escalation of the accident classification is provided by the staf f and management of the facility involved.

All these inputs are considered by the Emergency Response Team as they determine and continue to assess the situation, not only to take coriettive action, but to classify the accident properly in accordance with the definition given in Section 3.0.

1 In case of fire, in most cases, corrective action is immediately direc$.ed toward limitation and suppression of the fire.

In Soh. cases, firefighting 2

must be conducted with consider 6 tion for the spread of contamination or the hazard of potential criticality.

i While use and storage of flansnable materials are minimized insof ar as O

practical and inert atmospheres are provit'ed for all operations with highly reactive material, automatic fire sprinklers are installed in most areas.

(Fcr other safety considerations, they are not located in the hot cells.)

In adeittion to heat-actuated detector (HAD) ane products-of-coebcstion (POC) ser-i sors, the sprinkler system also provides a fire alarm to the Protective Ser-vices Control Center.

The alar., alerts firemen and also starts 0. call-in pro-

)

cedure for facility and radiation protection personnel.

If large amounts of l

water are used in suppressing a fire, containment and analysis of the water for radioactivity are performed.

(Most of the chemical and physical forms of i

SNH used at this facility are not soluble, and little transport by water wcu.d I

occur.)

Review of the exhaust, ventilation, and filtration systems is per-l forsed promptly, and any further potential for airborne releases is minimind by appropriate shutdown or restart of these various systems.

Needed actions

[

are identified by t5e reentry team and implemented at the direction of senior I

d i

staff members at the scene.

q l

Rl/R088-206 5-4

. D

In the case of a criticality incident, corrective actions involve assur-ing that all engineered safeguards systems are functioning appropriately.

Since the materials in use and the control procedures make an accidental cri-ticality a highly incredible event, only extremely unlikely combinations of material, moderator, and reflector could result in criticality.

Therefore, it is exper.ted that an accidental criticality would be quickly and permanently terminated.

Assessment of the accident would lead to a determination of the need to prevent recurrence by removal of reflector materials or disruption of the fuel configuration.

This would be done only after adequate planning and review.

In the event of a glovebox break, confinement is restored by use of tem-porary patches using plastic sheeting and tape.

This is done promptly by the reentry team or a follow-up crew.

Dispersed material is gathered and stored l

temporarily in criticality-safe containers.

Further recovery actions are developed by Operations and Radiation Safety staff members.

In most cases involving release of radioactive material, adequate protec-tion is achieved by preventing entry of unprotected personnel into contami-nated areas and by effecting an early clean-up operation.

Airborne resuspen-i slon is prevented by use of water sprays with coller. tion of run-of f water and vacuum cleaning of the most contaminated areas.

5.4 PR01EC11VE ACTIONS Protective actions are initiated by the Emergency Response Team, in most accident s.tuations, this involves evacuation of all personnel f rom the f acil-ity involved and keeping personnel in nearby facilities inside.

The suitabil-ity of emergency assemuly areas for c.ontinued use is jHged at the initial response and as the incident continues.

Indoor areas are available at both sites for isolation and care of large groups of people.

O Rl/R088-206 5-5

i For people in nearby buildings, sheltering is always the appropriate 1

k action for all of 1,he potential accidents.

All individuals are evacuated from a facility involved in an emergency.

If an emergency assembly area in use is downwind of a facility subject to a significant airborne release, or if the exposure rate exceeds a few mrem /h, the group is moved to another emer9sacy assembly area or to an indoor area.

Response personnel use suitable protective equipment and survey instru-e ments.

Contamination control points and surveys are established to prevent the spread of radioactivity during assessment and recovery operations.

5.4.1 Persnnnel Evacuation from Site and Accountability Personn" in a facility af fected by an accidental criticality evacuate immediately upon detection of the associated radiation by the Radiation Alarm System.

This activates a stren evacuation alarm and alerts the site Indus-trial Security Control Center.

An announcement is nude on the PA system in-forming all personnel of the existence of an accident condition in the af-fected building.

Personnel in nearby facilities are instructed to remain indooi4.

Certain facilities have procedures requiring shutdown of ventilation systems lo mini-mize the intake of any airborne radioactivity.

If a significant release of airborne radioactivity were anticipated, personnel could be relocated in near-by facilities or evacuated, if appropriate.

Since there is no reservoir of highly radioactive gas associated with a criticality accident, the release of airborne radioactivity would most likely be as a "puf f,"

and sheltering is more appropriate than evacuation.

Evacuation would become necessary in the case of a fire threatening to involve adjacent buildings.

All notifications can be ma4 by the PA system with the in-plant phone system as a backup.

PA notification is effective im-mediately.

Telephone notification to the few nearby buildings can be accom-plished in a few minutes.

O Rl/RD88-206 5-6

/~

On-site evacuation routes and reassembly areas have been established and posted.

All are within walking distance of the various f acilities at the site.

If it were desirable to relocate personnel to more distant areas, per-sonal autos would be used.

First-line supervisors are responsible for assuring that all personnel, visitors, and contract personnel in their units are accounted for at emergency assembly areas.

Missing persons are identified to Industrial Security.

Upon reentry of a facility by the response team, a search is made for any persons remaining in the facility.

Since there are only a few well-defined exits from the site, radiological monitoring of evacuees can be performed before personnel leave the site.

Facilities exist on the site for decontamination and medical care.

5.4.2 Use of Protective Equipment and Supplies All entries' into creas with unknown but potentially high levels of air-O borne radioactivity use self-contained breathing apparatus (SCBA).

Personnel using an SCBA are medically qualified and trained in its use.

Face-piece fit tests are performed periodically to assure adequate protection.

Initial reentry utilizes a one-piece disposable coverall with integral shoe covers and hood.

This is supplemented by rubber gloves, canvas shoe covers, and a full-face respirator with 30 min air supply.

Adequate supplies of personal protective clothing and equipment are kept in the emergency response van 6t Santa Susana.

Additional supplies are avall-able in several facilities at the site.

On-site distribution can be effected by truck, van, scooter, or handcart, as appropriate.

Depending upon conditions found by ;he reentry team, other apprupriate protective clothing, including underwear, gonad shields, and waterpioof overalls, are available, i O l

Rl/R088-2U6 5-7

o Respiratory protection is selected f rom the following types available, (M

according to the needed protection factors:

Protection Type.

Factor e

Half face with Type H ultrafilter 10 Full face with Type H ultrafilter 50 full face with supplied air, pressure demand 2,000 Full face with supplicc air, continuous flow 2,000 Full face, SCBA, pressure demand 10,000+

Isolation and area access control are provided by the arrangement of the af fected facility in most cases.

Additional control is established at any severely contaminated areas by use of temporary barriers and step-of f lines.

Shadow shields could be constructed of lead blocks, available on site, when necessary, for local shielding.

Return of an area to normal use requires satisfaction of the contamina-

[

tion and radiation exposure limits established in Special Nuclear Materials C

License SNM-21, as shown in Tables 5-1, 5-2, and 5-3.

Responsibility for decentamination to these limits is assigned to the operations group responsible for the f acility.

Compliance is determined by Radiation and Nuclear Safety.

5.5 EXPOSURE CONTROL IN RADIOLOGICAL CON 11NGENCIES Radiological exposures to emergency workars are controlled by use of direct-reading pocket dosimeters and limitation of stay times.

5.5.1 Emergency Exposure Control Program Exposures to individuals in eraergencies are limited according to the fol-lowing criteria previously established and approved by NRC for activities RI/RD88-206 5-8 i

I I

Table 5-1.

Smear Test Surface Contamination Limits (m) and Action Guides j

v Limits 2

(dpm/100 cm )

Restricted Areas Unencapsulated Type of R/AMate{ial Other Contaminant Areas Areas Unrestricted High-RCG*

440,000 4,400 1,000 Low-RCG, 20,000 700 200 Low-SA*

Low-RCG, 440 20 MOL**

High-SA ActionGuidgs (dpm/100 cm )

High-RCG*

5,0005 6005 1N Low-RCG, 200 100 20 Low-SA*

F Low-RCG, MOL**

MOL**

High-SA

RCG:

Radioactivity Concentration Guide in air SA: Specific activity tFor this table Unencapsulated R/A Material Areas include only those areas in which work is being conducted with unencapsu-lated radioactive materials.

$1ncrease by a factor of 3 for tritium

- Minimum detection limit, using standard smear techniques and appropriate counting techniques I

f f

O RI/R088-206 5-9 l

Table 5-2.

Radionuclide Grouping Criteria g

RCGing)ir RCGinAjr D

Type of Contaminant (vg/cm (pC1/cm )

-10 High-RCG Not specified

> x 10

-8

-10 Low-RCG, Low-SA

). x 10

< x 10 Low-RCG, High-SA

< x 10-8

< x 10-10 Table 5-3.

Surve Instrument Surface Contamina: tion Limits Type of Contaminant Limit High-RCG 0.2 avg.,1.0 max mrad /hr at surface Low-RCG, low-SA 5 avg., 25 max dpm/cm

Low-RCG, High-SA MOLT

This limit is applicable to uranium enriched in U-235 and to natural and depleted uranium which has been subjected to a recent melting process.

For other natural and depleted uranium, the limit for high-RCG contaminants applies.

tMinimum detection lim t (using available instruments such as pancake G-M detectors and alpha scintillators),

t licensed under Special Nuclear Materials License No. SNM-21.

Thus, it is pre-sumed that exposures received under such conditions will not be judged to be in noncompliace with regulatory requirements.

5. 5.1.1 Evoosure Guidelines Reentry shall be conducted to assess conditions existing after an inci-dent, to locate and ratrieve survivors, if possible, and to identify the cause of the incident.

The following* criteria will be applied to reentry cperations:

1.

Initial Reentry Dose limited to 200 mrem Exposure rate limited to I rem /h

O Rl/ROFJ-206 5 10

If these conditions occur during the initial reentry, the team will

(

withdraw, and plans will be developed for follow-up reentry opera-(

tions.

2.

Follow-Up Reentry Dose limited to occupational limits (1.25 rem to 3 rem in current quarter) consistent with ALARA.

3.

Recovery--Lifesaving Dose limited to 75 rem, for volunteers only, approved by Radiation and Nuclear Safety at the time.

(Risk must be judged against the prcbability of success.)

4.

Recovery--Deceased Victims Dose limited to occupational limits (1.25 rem to 3 rem in current quarter) consistent with ALARA.

5.

Recovery--Protection of Health and Property a.

Minor hazard--occupational dose limits b.

Moderate hazard--12 rem total for current year c.

Severe hazard--25 rem for the operation d.

Extreme hazard--75 rem, for volunteers only, approved by Radi-ation Safety at the time.

Exposures are monitored by means of direct-reading pocket dosimeters sub-ject to individual control and to review by Radiation and Nuclear Safety staff l

members at the scene.

Limitation of doses is established by use of stay-time limits based on exposure rates, measured or estimated, and dosimeter readings.

All employees assigned to work with radioactive materials and assigned to re-l entry and recovery teams are qualified to receive up to 3 rem in a current quarter.

All exposures are subject to the review and approval of the senior member of Radiation and Nuclear Safety available at the scene or by telephone.

I i

Exposure to radiation during recovery operations, including provision of first aid, performing personnel decontamination, providing ambulance service, and providing medical treatment services, are limited to the values estab-i lished in 10 CFR 20.101 for occupational exposures (including any outside of a l

restricted area) and 10 CFR 20.105 for nonoccupational exposures.

l Rl/RD88-206 5-11

m 5.5.1.2 Radiation Protection Procram Exposures during reentry and recovery operations are limited by the responsible member of Radiation and Nuclear Safety at the scene according to the preceding guidelines.

Actions requirie, axposure limits of up to 75 rem are permitted to knowledgeable volunteers only, upon approval by the Manager.

l Radiation and Nuclear Safety; the Director, Health, Safety & Environment; or their delegated alternates.

This approval may be obtained by telephone, based upon consultation with appropriate personnel from the groups involved.

Strin-gent controls would be applied to tasks that are estimated to approach 75 rem.

5.5.l ' 3 Monitorina Internal depositions of radioactive material are determined by full implementation of the bioassay program, as indicated by contamination survey results and measurements of airborne radioactivity.

Analysis of urine and fecal samples is performed by U.S. Testing, Inc., in Richland, Washington, with a normal response time of approximately 2 weeks.

Screening analysis for gamma 'mitters could be accomplished by use of a high-resolution gamma spec-trometry system at SSFL.

This system uses a high-purity germanium detector and provides qualita-tive (radionuclide identification) and semiquantitative measurement of person-nel internal or external contamination.

.115 type of usage has been demon-strated by trials on individuals receiving diagnostic and therapeutic radio-phannaceuticals (1-131).

I In-vivo measurecent could be performed at several schools and hospitals with nuclear medicine f acilities in the Los Angeles area.

For example, an in-vivo counting system, calibrated for I-131 and other nuclides, is available at UCLA, approximately 25 miles away.

lO RI/RDB8-206 5-12 4

O An adequate supply of direct-reading pocket dosimeters in various expos-ure ranges is available.

Special film badges are assigned to all personnel 1

involved in significant exposures, in addition to normally worn personal film badges.

Inhalation exposures are assessed on the basis of nasal smears, lapel air samplers, fixed air samples, continuous air monitors, and other samples of airborne radioactivity.

All data are kept as part of the permanent personnel exposure record system.

5.5.2 Decontamination of Personnel Decontamination of personnel is required for any detectable contamina-tion.

No individual, clcthing, or equipment is released without special pre-cautions unless levels of contamination are below those shown in Table 5-4, which are required by Condition 23 of License SNM-21, that establishes Annex B, "Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct. Source, or Special Nuclear Material.

Decontamination of personnel is accomplished by washing, using nonabra-sive soap, at a sink that drains into a radioactive liquid hold-up tank, and spot cleaning by use of liquid window cleaner and paper towels, with the towels kept for disposal as radioactive waste.

For dry contaminants on hair-free areas, sticky tape is used.

For difficult cases, potassium permanganate/

sodium bisulfite is used. More severe cases require medical supervision, t

Decontamination of wounds is performed by medical personnel.

This is l

performed at the medical station or at the fully equipped medical decontami-nation facility at Santa Susana.

Adegiate supplies of workmen's coveralls are available if it is necessary to remove and hold personnel clothing for decontamination, radiation measure-I ments, and/or disposal.

Contaminated equipment, supplies, and instre ients are decontaminsted in suitable laboratory facilities.

RI/Ruu8-206 5-13

[N O

V Table 5-4.

Acceptele Surface Contamination Levels herage,c.f Maxisemp,d.f 3,,,,gje.e,f b

b aluclidesi u-nat, U-235, U-238, and 5,000 dem a/100 anz 15,000 dpm a/100 cm2 i,ogo d,,ofico cm2 associated decay products 2

2 2

Transuranics Ra-226, Ra-228, 100 dpm/100 an 300 dpm/100 an 20 dpm/100 aW Th-230 Th-228, Pa-231 Ac-227, 1-125, 1-129 2

I 2

th-nat, Th-232, Sr-90, 1,000 dyn/100 an 3,000 dpm/100 oW 200 e5m/100 cus Ra-223, Ra-224, U-232, 1-126, 1-131, 1-133 2

2 2

Beta emitters (nuclides 5,000 dpney/100 aw 15,000 dpm By/100 aW

000 dpm By/100 as with ay modes other than alpha emission or spontaneous fission) except Sr-90 and others noted above h

'hfiere surface contmination by both alpha-and beta-ganna-emitting nuclides exists, the limits estelished for g

for alpha-and beta-gasuna-emitting nuclides should apply indepsedently.

hg bAs used in this table, d(un (disintegrations per minute) means the rate of emission by radioactive material as 4

as detemined by correcting the counters per minute observed by an appropriate detector for background, effic.

iency, and geometric factors associated with the instrumentation.

c3 e

CMeasurements of average contminant should not be averaged over more than I sguare me'er. For objects of less of less surface area, the average should be derived for each such object.

d 2

ihe maximen contmination level applies to an area of not more than 100 on.

'The mount of removele radioactive material per 100 an2 of surface area should be detesinined by wiping that area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instrument of krown efficiency. 1sien remov ele cont ainatica on ob* ts of less surface area is detemined, the pertinent levels should be reduced proportionally and the entire ace should be wiped.

IThe average and maximum radiation levels associated sith surface containation resulting from beta-gasuna emitters should not exceed 0.2 mrad /h at 1 on and 1.0 mrad /h at 1 an, respectively, measured through not more than 7 milligrms per square centimetee of total absorber.

5125Y/ paw

5.6 MEDICAL TRANSPORTAf!0N Although a well-equipped, company-owned ambulance operated by employee emergency medical technicians (EMis) at SSFL is available, triage considera-tions must dictate the mode of transporting the injured.

Victims of crush injuries (most common in earthquties), falls of over 20 f t, and penetrating wounds, for example, must be conside.*ed to have life-threatening internal bleeding with time the most important factor.

In such situations, a heli-copter

ambulance with a trauma surgeon and surgical nurse on board, as UCLA's Med-Star does, could be life saving.

As mentioned above, LAFD paramedic ambu-lances and two local commercial ambulance companies are prepared to transport contaminated patients, in an area-wide disaster with comunity services overloaded, transporta-tion of the seriously injured may have to be by company or employee vehicles, provided roads are passabic.

Prior to transportation, unless medical triage considerations rule other-wise, personnel may be decontaminated using on-site facilities as described in Section 5.5.2.

When ambulance / helicopter transport of contaminated personnel is necessary, a health physicist must accompany or meet the ambulance at the hospital.

The available on-site medical treatment facilities and capabilities are described in Section 5.7.

The arrangements with the local hospital for ac-cepting contaminated patients are also described in Section 5.7.

5.7 ME0! CAL TREA1 MENT Initial encounter medical services, including triage, primary diagnostic evaluation and treatment, and cardiopulmonary resuscitation, as needed, are provided by the Rocketdyne Medical Director and staf f, utilizing the medical facilities at the Santa Susana Field Laboratories (SSFL), the Canoga and I

i O

RI/RD88-206 5-15

/' N De Soto medical stations.

Medical facilities at SSFL include a decontamina-tion treatment trailer located near the SSFL medical station and the Indus-trial Security control Center.

The medical facilities at Rocketdyne include well-equipped examination and treatment areas to handle emergencies.

The Medical Director, Physician's Assistant, and Occupational Health h

Nurse at SSFL are graduates of Oak Ridge Associated Universities (0RAU) Radi-ation Emergency Action Center Training Site (REAC/TS) courses on radiation accident management under the auspices of the Department of Energy (00E).

In addition, both the Medical Director and Director of Nurses of the Emergency Room of Humana Hospital West Hills, the nearest hospital to SSFL, are REAC/IS trained.

The primary supporting hospital emergency room for diagnosis and treat-ment of non-Level-1 trauma cases, that is, severely injured, including con-taminated patients, is Humana Hospital West Hills, 7300 Medical Center Drive, O'

Canoga Park, (818) 884-7060.

In addition, a partial list of area hospitals that are qualified in the treatment of radiation accident victims is included in Table 5-5.

5.7.1 Medical Triage Triage involves life and death medical decisions under "battle-field" conditions whereby victims are categorized and their treatment prioritized by survivability.

Tags are assigned to wounded victims accordingly:

Red:

Emergency--treatment needs immediate for survival, needs trauma Level-1 care.

Yellow:

brgent--treatment may be delayed for survival.

Green:

Wounds not life threatening, may wait until Red and Yellow cases stabilized.

Black:

Fatality--need body bag.

O RI/R088-206 5-16

Table 5-5.

A Partlai List of Area Hospitals Capable of Radiation Accident Management

\\

(Sheet 1 of 2)

Institutions able to handle radiation accident victims, including designation of appropriate isolation facilities and procedures for handling and removal of waste, and the maintenance of trained personnel:

1.

HAWTHORNE COMMUNITY HOSPITAL 9.

NORTHRIDGE HOSPITAL FOUNDATION 11711 Grevillea Avenue 18300 Roscoe Boulevard Hawthorne, California 90250 Northridge, California 91324 973-1711 885-8500 2.

HARBOR GENERAL HOSPITAL 10.

PACIFIC HOSPITAL OF LONG BFACH 1000 West Carson Stress 2776 Pacific Avenue Torrance, California 90509 Long Beach, California 90801 328-2380 595-1911 3.

HUNTINGTON MEMORIAL HOSPITAL 11.

SAINT JOSEPH MEDICAL CENTER 100 Congress Street 501 South Buena Vista Street Pasadena, California 91105 Burbank, California 91505 796-0371 843-5111 4.

KAISER FOUNDATION HOSPITAL --

12. SANTA MONICA HOSPITAL LOS ANGELES ME0! CAL CENTER 4867 Sunset Boulevard 1225 - 15th Street Los Angeles, California 90027 Santa Monica, California 90404 O

667-4011 451-1511 5.

KAISER FOUNDATION HOSPITAL--

13. UNIVERSITY OF CALIFORNIA, LOS WEST LOS ANGELES ANGELES, HOSPITAL 6041 Cadillac Avenue 10833 LeConte Avenue Los Angeles, California 90034 Los Angeles, California 90024 857-2000 825-7271 6.

LOS ANGELES NEW HOSPITAL 14.

VALLEY HOSPITAL 1177 South Beverly Drive 14500 Sherman Circle Los Angeles, California 90035 Van Nuys, California 91405 553-5155 997-0101 7.

Lf,NG BEACH COMMUNITY HOSPITAL 15.

HUMANA HOSPITAL WEST HILLS 1720 Termino Avenue 1300 Medical Center Drive Long Beach, California 90801 Canoga Park, California 91307 597-6b55 884-7060 8.

MARTIN LUTHER KING, JR.

16.

NU-MED REGIONAL MEDICAL CENTER HOSPITAL 22141 Roscot Boulevard 12021 Wilmington Avenue Canoga Park, California 91304 Los Angeles, California 90059 340-0580 639-8550 R!/R088-206 5-17

1 l

I i

y l

Table 5-5.

A Partial List of Area Hospitals Capable of Radiation Accident Management l

(Sheet 2 of 2) l l'

i 17.

LOS ANGELES COUNTY--USC t

j

' GENERAL HOSPITAL i

i 1200 North State Street Los Angeles, California 90033 226-2622 l

18. MEMORIAL MEDICAL CENTER OF I

LONG BEACH 2001 Atlantic Avenue i

Long Beach, California 90801 595-2311 l

t i

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Rl/RD88-206 1

i 5-18 1

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c..

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Triage decisions are to be made by medical personnel; the doctor, PA-C, nurses, and may have to be delegated with the aid of two-way radios to protective services officers, Health & Safety personnel, Health Physicists, or volunteers.

5124Y/ paw O

O Rl/R088-206 5-19

/~'S 6.0 EQUIPMENT AND FACILITIES b

6.1 CON 1ROL POINIS The primary control point is the Industrial Security Control Centar at the Santa Susana site.

Secondary control points are located at preestablished Emergency Assembly Areas (EAAs) as near the incident location as possible.

Two secondary control points are established at EAA 3 and EAA 5, as shown in Figure 6-1, at the Santa Susana site.

These same areas are used for assem-bly areas for personnel collection af ter evacuation.

They are sufficiently removed from areas of possible radiological accidents to be minimelly affected.

6.2 COMMUNICATIONS EQUIPMENT During the early stages of an emergency, until conditions are stabilized and plant restoration is initiated, the Emergency Response Team has access to the Industrial Security Department radio net.

This allows either simplex or

\\

repeater operation between the control' point, any Industrial Security person-nel, and Industrial Security Communications Centers.

This provides a communi-cation link to any Rocketdyne staff member or external agencies.

Telephones are provided at the primary control point and at secondary

~

I control point EAA 5; additionally, telephones are available in adjoining or nearby facilities.

Communication requirements for all subsequent recovery needs are satis-fled with commercial telephone service.

6.3 FACILITY FOR ASSESSMENT TEAM Facilities to be used by the Emergency Response Team are designated on an as-needed and case-by-case basis and determined by several f actors including triage requirements, adequacy of facility, personnel, supplies, communications

!O Rl/RD88-206 6-1

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q equipment (telephone / radios), and the availability of appropriate medical (V

evacuation equipment (ambulances /Medi-Vac helicopters).

In the extreme, a field medical team would be established on the scene.

6.4 ON-SITE MEDICAL FACILITIES The location and capabilities of the on-site medical f acilities for decontamination and primary medical treatment were covered in Section 5.7.

The Rocketdyne medical stations are equipped and supplied to provide initial treatment in emergency and less urgent level medical problems as required for diagnosis and treatment and include EKG computers, pulmonary function testing units, X-ray (Canoga only), examining tables, beds, surgical instruments, sphygmomanometers, pha rmac euti':als,

analgesics,

bandages, splints, etc.

In addition, as a 00E physician, the Medical Director has on hand, through the assistance of Oak Ridge Associated Universities, the chelation gxb agents trisodium ca'lcium pentamil and trisodium zinc diethylene-triaminepenta-acetate (ZnDTPA), new drugs limited by Federal law to investigational use for the treatment of human contamination with plutonium.

6.5 EMERGENCY MONiiORING EQUIPMENT A variety of monitoring equipment is kept in use throughout the Santa Susana f acilities.

These provide the capability to monitor personnel expo-sures under accident and recovery conditions, area monitoring for radiation and contamination, developing a basis for estimates of releases to the environment, and to measure and record meteorological conditions.

The various instruments are listed and described briefly in Tables 6-1 through 6-5.

V Rl/RD88-206 6-3

Table 6-1.

Personnel Monitoring Instruments Detector Calibration Alare Set Equipment Type Sensitivity Range Pouer Freguency Alarm Point Ludlun Model 12 G-M pancake 0.2 cpa/dpa 0-500 Kcyn Battery 13 weeks Tech Assoc pug-1 G4 pancake 0.2 cyn/dpm 0-50 Kcan Battery 13 useks Ludlun Model 12 Alpha scint 0.2 cpm /dyn 0-500 Kcpm Sattery 13 weeks

[h Ludlun Model 177 G4 pancake 0.2 com/dyn 0-500 Kcpm 110 V 60 Mz 26 weeks Local variable 4

with battery horn + light o"

Ludlun Model 177 Alpha scint 0.2 cyn/dyn 0-500 Kcpn 110 V 60 Hz 26 useks Local Variable with battery horn + light 5125Y/pau

p s

C' idle 6-2.

Area Monitoring lasiruments Detector Calibration Alare Equipment Type Sensitivity Itange Power Frequency Alarm set point Ludlun Model 12 G4 pancde 0.2 cyn/dyn 0-500 Kcpm sattery 13 weets Tech Assoc PUG-1 G 4 pancd e 0.2 cr dpm 0-50 Kcyn Settery 13 weeks e

Ludlun Model 12 Alpha scint 0.2 com/dyn 0-500 Ecyn Battery 13 weeks Ludlus Model 177 G 4 pancake 0.2 cpa/dym 0-500 Kcyn 110 V 60 Hz 26 weeks Local Varidle with battery horn + light g

Ledian MLJe1 177 Alpha scint 0.2 cpa/dyn 0-500 Kcre 110 V 60 Hz 26 weeks Local Varldle Tg with battery here + light Reuter-Stokes High-pressure few wit /h I-500 wit /h 110 V 60 Hz 6 months Strip chart rec N g

RS-Ill ion chamber with battery MMC GA-2T Mal (TL)

I adt/h 0.1-1,000 aft /h 110 V 60 Hz 1 months Local light, 20 mA/h sires, control Tracer 14 14P-1 G4.

I aft /h 1-1,000 adt/h 110 V 60 Hz 3 aseths Local light.

20 um/h stree, control Ludlun Model 125 Nel (II) few pit /h 0-3,000 pit /h Battery 13 ameks 5125Y/ paw i

b

Os U

,s a

Table 6-3.

Instruments for Measuring Releases to Enviresumet Detector Calibration Alare Equipment Type Sensitivity Range Pouer Frequency Alare 5et Point Acuter-Stokes High-pressere few yR/h 1-500 >R/h 110 V 60 Nr 6 months RS-111 los chanter with battery Strip chart recorder victoreen CaF :Mn few adt I am-1,000 R nune prior to 2

thernoluminescent bulb (or greater) ese Requires readoet 3

dosimeters s

fh Staples high-Particulate fewfCiM 110 V 60 Hz 6 months y

volone air filter or 24 V de seguires analysis e

g sas,1er e

Hi-Q high-Particulate few fCl/m3 12 V dc 6 mot.ths volume air filter Requires analysis sampler Environmental Particulate few fCi M 110 V 60 Mz air samling filter neguires analysis l

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Instruments available for monitoring radioactive contamination and suking

(T radiation surveys include the following:

Contamination monitors, portable:

Emergency use--4 Routine use--24 Contamination monitors, ac-powered:

Emergency use--

Routine use--26 Radiation survey instruments, portable:

Emergency--2 Routine use--63 These instruments for both emergency and routine use are maintained, serviced, and calibrated at quarterly intervals for the portable equipment and semi-annually for ac-powered equipment.

O 5124Y/ paw l

l I

i l

7 O

Rl/R083-206 6-8

p 7.0 MAINTENANCC OF RADIOLOGICAL CONTINGENCY PREPARE 0 NESS CAPABILITY U

7.1 WRITTEN PROCEDURES The written procedures prepared for the guidance of the Emergency Re-sponse Team are subject to an independent internal review as part of manage-ment approval.

They are periodically tested for adequacy during the periodic tests and drills (see 7.3 below).

If, as a result of the periodic tests, any weaknesses or omissions in the procedures are noted, the procedures will be corrected as required and again independently reviewed and approved by manage-ment.

The responsibility for updating or changing the emergency plans for management approval rests with the Emergency Coordinator, reporting to the Director of Industrial Security.

7.2 TRAINING Personnel responsible for emergency at tions at the direction of the Emergency Response Team leader consist of Operations personnel at f acilities O

where licensed activities are underway, health physicists, and Protective Ser-vices of ficers.

When personnel are assigned to these responsibilities for i

radiological emergency service, they are given the following training:

Basic training in self-contained breathing apparatus Protective clothing suit-up

{

Radiation safety l

e Criticality safety l

e t.icensed facilities familiar 1 ration i

Fire extinguishment e

Radiological Contingency Plan-accident classification, notification, procedures, responsibilities, and comunication equipment Retraining is required semiannually on:

Self-contained breathing apparatus Protect Ive clothing suit-up e

Licensed facilities familiarization

[

e I

O RI/R088-206 7-1 l

t

Retraining is required annually on:

.,(

)

Radiation safety Criticality safety Fire extinguishment e

Radiological Emergency Plan Exceptions to the above training and retraining requirements are made in the case of personnel whose normal duties or knowledge and experience involve these subjects or topics.

The Technical and Skills Development Department is responsible for pre-paring the training syllabus, preparing and giving the training, scheduling into the training courses personnel who are identified by management as re-quiring the training, maintaining training records, and notifying management when retraining is required for personnel in compliance with this plan.

The manager responsible for each f acility involved with licensed activi-ties, the Dliector of HS&E, and the Director of Industrial Security are re-

)

sponsible for providing the names of those personnel who require thl's training to Technical and skills Development.

The Emergency Coordinator monitors these courses as they are given and makes recommendations to his management as to course content.

The training program is audited to this plan on an annual basis by Quality Assurance.

Quality Assurance reports these findings to management and the Emergency Coordinator.

7.3 1ES15 AND ORILLS Annual exercises are carried out by simulating an emergency at Build-ing 020 at Santa Susana.

These exercises involve the Emergency Response Team and alternates and, as necessary, depending on the emergency being simulated, the support personnel that would be expected to be called in to serve the team.

This would include those off-site personnel, such as the Ventura County i

v RI/R088-206 1-2

Sherif f's Department, the Ventura County Fire Department, ambulance services, d

and off-site medical facilities as appropriate to the exercise either as par-ticipants or as observers.

Scenarios for these exercises are prepared by the Radiation and Nuclear Safety staf f and the Emergency Coordinator (reference 4.2.2.3) to provide as meaningful an exercise as possible consistent with the overall safety of the participants and economic considerations.

Certain observers are designated to assist in the response evaluation of personnel, instrumentation, safety system, and equipment.

7.4 REVIEW AND UPDATING OF THE PLAN AND PROCEDURES i

Emergency plans and procedures will be updated as needed.

These needs will be identified as a result of continual management review and as a reselt j

l of reviews of the reports provided by observers of the tests and drills. A1) l critiques of actual emergencies will include consideration of any possible r

changes in the emergency plans, engineered safeguards system, process 4

I l

procedures, process system, plant organization, and other matters bearing directly on safety questions.

)

7.5 MAINTENANCE AND INVENTORY OF RADIOLOGICAL EMERGENCY EQUIPMENT, i

1 INSTRUMENTATION, AND SUPPLIES i

i The inventory of emergency supplies at the Santa Susana site is given 'n r

Table 1-1.

The inventory is checked against this list by Industrial Security weekly.

The radiological instrumentation and self-contained breathing appar-l atus are maintained to assure working order and calibration by the Calibra-tion, Recall, and Inventory System (CRIS).

This system provides notification that an instrument is due for recall for inspection, calibration, or ad,1ust-

[

ment. Work that is not performed as schsduled is printed out on a delinquency i

report for management attention, f

5124Y/ paw O

i R1/R088-206 7-3 L

l

Table 7-1.

Emer,ency Supplies at Santa Susana Field Laboratory Site

(

)

' s' Quan-Quan-tity item tity Item 1

Roll plastic 2

Roll, Tuck Tape (green) 1 Axe Extra 5-gal Plastic Bags 1

Bar Extra Surgeon's Gloves 1

Shovel Extra (Bunny Suit) RA Clothing 2

Grey-Lite Suits 2

Fire Blankets 4

Mask MSA 1

Hand Light i

8 Pali

Leggings (Sodium Fire) 1 Sprinkler Shutoff 1

Emergikit !! Resuscitator 8

MSA Goggles i

1 Spare Oxygen Bottle 1

Kwik Cold Ice Pack i

large ',oll Aluminum Foil 1

Box Dust Respirators 6

Spare MSA Bottles 3'

Black and Yellow Lines 6

Extra Far Com Ear Plugs 2

Life Liners 6

Complete 3ets Red-Line Protec-2 Safety Harnesses tive Clothing 1

Road Block Sign 6

Flame-Resistant Coveralls 6

Hard Hats with Shields (Sodium Fires) 1 Ludlum Mod'el 12 Alpha Counter 3

' Pair Leather Welder's Gloves 6

Pair PVC Gloves 1

Ludlum Model 12 Beta-Gama Counter 1

15-lb CO Extinguisher 2

6 Pocket Dosimeters 1

30-15 Na, Exdnguisher 1

Bendix Dosimeter Charger 4

MSA Packs with Regulators (Hodel 906-2) 1 MSA First Aid Kit i

Eberline Model R0-2 (0-5 R/h 3

Spare Radio Cases Gamma and X-ray) 1 1

3 Ear com 1

Eberline Model R0-2A (1-50 R/h Gama and X-ray) 2 Portable Electric Air Sampler and Filters (l?-V) (Gun Box) 1 First Aid Box (727) t

[D

(

)

v RI/R088-206 7-4 l

8.0 RECORDS AND REPORTS g ]

8.1 RECORDS OF INCIDENTS Incident reports for the record will be made for "unusual" events and the more serious event classifications (see 3.0).

The preparation of these reports will be the responsibility of the facility / operational management of the involved f acility.

The report will include the cause of the incident, psrsonnel and/or equipment involved, extent of injury and/or damage, correc-tive actions taken during the emergency, recovery steps taken to return to operation, and the actions taken to prevent recurrence.

Participation and site and of f-site support assistance, with recommendations evaluati (if any) i

_provements, will be included.

The records will be maintained and retained by the Director, Nuclear Safety & Licensing, as long as the facility / operation is active in licensed programs.

8.2 RECORDS OF FREPAREONESS ASSURANCE 3

sj Records will be maintained for all drills and tests of the plan.

These records will include the reports of observers and their recommendation for improvement.

These records, along with inventory records of emergency sup-plies and agreement letters for off-site support organizations, are mair.tained by Industrial Security for a period of 2 years (ra*erence 4.2.2.3).

Mainte-nance, testing, and celibretion records on equipment required for emergency response (whether stored or in active use) are maintained by the CRIS system as described in Subsection 7.5.

Records of training, as required by Subsection 7.2, are maintained by Technical and Skills Development.

These records are retained on site for a minimum of 2 years beyond which time they are submitted to long-term storage (75 years) off site.

Rl/RD88-206 8-1 I

8.3 REPORTING ARJANGEMENTS O

The notifications required for each of the four classes of events are specified in Table 3-3 for the on-site plan.

This includes Rocketdyne manage-ment and the hRC.

Section 4.0 describes the arrangements that have been made for notification of other agencies required for support.

Subsection 4.2.2.5 describes the duties of the Comunications representative on the Emergency Response Team for comunicating to corporate management and members of the pub 1'c and press regarding the status of on-site emergencies.

5124Y/ paw O

R!/R088-20f 8-2

9.0 RECOVERY Recovery from any radiological event is under the dire: tion of the Emer-gency Response Team leader unless earlier release to the facility management is formally agreed to between the responsible parties end is approved by management.

Recovery in this chapter is defined to begin after the event has ended, when there are no further considerations of any possible life-saving actions, and when conditions have stabilized suf ficiently to allow careful recovery planning.

however, recovery planning will be influenced if there is an indication that additional risks are likely because certain decontamination steps have not been taken to limit or reduce exposure risks, or safety features need to be reestablished to provide needed protection, or structural defects corrected to assure f acility integrity.

The reentry criteria stated below will assume that one or more of these considerations exist.

The criteria for entry into tha area during plant restoration where none of these urgent conditions exist will be controlled by the Radiation and Industrial Safety policies followed at Rocketdyne during normal operations.

l 9.1 REENTRY 9.1.1 Radiological Criteria During recovery, the radiological criteria imposed will assure that the exposure limits will not exceed the exposure limits established by 10 CFR 20.

9.1.2 Indusicial Safety Criteria Rocketdyne policies for industrial safety practices will be followed.

O R1/R088-206 9 -1

(3 9.2 PLANT RESTORATION U

Plant restoration is the continuation of the recovery process af ter any of the immediate safety items discussed in Subsection 9.0 above have been resolved.

The radiological control criteria and indus'.cial safety criteria to be enforced during restoration will be those that are imposed for normal plant operation.

Restoration will include glans to assure that the facility will meet pre-event safety standards for containment and that required safety features are functional, including r?diaticn and criticality alarms.

Additionally, the plans will require that the emergency equipment desig-nated'in Section 5.0 has been returned to service.

9.3 RESUMPTION OF OPERATIONS The approvals for restart have been included in Table 3-3.

These ap-s provals 'will require that plant restoration plans have been completed and that the action items 'resulting from the investigation of the event have been resolved.

5124Y/ paw O

I Rl/R088-206 9-2 l

1

r____

APPENDIX A V

DESCRIPTION OF IMPLEMENTING PROCEDURES FOR EMERGENCIES Rocketdyne Operating Policy R0P M-501, Ens gency Incidents, governs the basic responsibilities fur reporting and handling emergencies.

This policy defines an incident as au occurrence which involved significant, actual, or potential prope, ty damage, property loss, personal injuries, fatalities, environmental pollution, or disruption of work schedules.

Examples include, but are not limited to, fires, floods, explosions, thef ts, earthquaket, bomb threats, $rbotage, civil disturbances, releases of radioactive or toxic sub-stances, and xcessive personnel radiation exposure.

In addition to the specific policy cited above, there are other operating policies and procedures pertinent to emergency planning and action.

These documents include the following:

R0P M-500 Rocketdyne Safety Program R0P M-502 Nuclear Criticality Control V

ROF M-504 Radioactive Material and Ionizing Radiation R0P M-508 Areas Requiring Special Safety Precautions (Radiological and Nonradiological)

R0P M 509 Unattended Tests R0P M-511 Nuclear Material Control and Safeguards Manageraent R0P M-512 Protective Garments and Safety Equipment HS&E G-01 Radioactive Materials and Ionizing Radiation Standard operating policies and procedures require that all emergencies be immediately reported to the continuously manned Industrial Security Control Center.

The Control Center, upon receiving an incident report or an alarm annunciation f rom detection or suppression alarm systems, will immediately i

notify and initiate action, as warranted by the nature of the incident, by i

emergency response functions which include:

f Industrial Security police and fire protection functions Health, Safety & Environment RI/R088-206 A-1

L Af fected facMy operating department and personnel

'V Medical Facilities and Industria' Engineering Comunications Operating procedural res90nsibilities are further implemented by the requirement for specific emergency action plans and procedures from emergency response functions and specific operating departments.

Specifically, with respect to a radiologicel eh.ergency, the first cate-t gorization of an incident requires the Industrial Security Control Center to immediately initiate notification to the radiological Emergency Response Team.

The response team consists of those members described in Section 4.2.

Upon arrival, the Emergency' Response Team evaluates the initial categori-zation and confirms, upgrades, or downgrades the incident classification.

The evaluation process continues until the event is terminated.

Industrial Secu-rity shall make those notifications as directed by the response team leader in accordance with the requirements noted in Table 3-3.

The procedures, responsibilities, and action plans for emergency response functions in support or under the direction of the Emergency Response Team are detailed in the Rocketdyne Master Emergency Plan, which has been previously approved as part of the license, and has been supplemented with this Radio-l logical Contingency Plan.

In addition, as appropriate to emergency planning, there is a specific emergency plan on file for the Hot, Cell, Building 020.

5124Y/ paw O

Rl/R088-206 A-2

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