ML19270H317
| ML19270H317 | |
| Person / Time | |
|---|---|
| Site: | 07001201 |
| Issue date: | 04/30/1979 |
| From: | Zeff D BABCOCK & WILCOX CO. |
| To: | Rouse L NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| NUDOCS 7906260185 | |
| Download: ML19270H317 (40) | |
Text
/
Y pp8-Babcock &Wilcox g,,,,c.,,,,,,,c,,,,
4 P.O. Bex 1260, Lynchburg, Va. 24505 Telephone: (804) 384-5111 April 30, 1979 United States Nuclear Regulatory Commission Office of Nuclear >!aterial Safety S Safeguards Division of Fuel Cycle and Material Safety Washington, D. C. 20555 A'ITENTION:
Mr. L. C. Rouse, Chief Fuel Processing and Fabrication Branch o
REFERENCES:
(1) SNM-1168, Docket g70 '
(2) Memo from M. A. GTora to L. C. Rouse, Amendment Request 78-4, 10/13/78 (3) Memo from W. T. Crow to D.'W.
Zeff, 12/17/78
SUBJECT:
Amendment Request 78-4 Gentlemen:
In response to your questions and comments in Reference 3 above con-cerning Amendment Request 73-4, we have revised certain sections of the submittal as noted in the index and are resubmitting the entire package for your review. The attached revision index outlines the changes that would result upon approval of the amendment.
Contamination " action levels" have been revisad to be more in accord with previously approved license limits and proposed Regulatory Guide S.24 (Health Physics Surveys During Enriched Uranium-235 Processing and Fuel Fabrication).
In addition, clarifying statements are made regarding the intent of the action levels.
We appreciate your consideration in this matter.
Please feel free to contact me should questions arise in the course of your review.
Sincerely, BABC0CK S WILCOX COMPANY COMMERCIAL NUCLEAR FUEL PIM T I,
l' h Ih:
d.-T.Zeff,Ma,na,.vi w
ger Safety, Licensing, and Safeguards DWZ:cmm 2318 153 Attachments 7 9 Q g g g 0\\W WM w/o attachments 3
The Bacccck & Micex Ccmpany l E :achsned 1867
SNM-1168 DOCVsET 70-1201 REVISIO:1 IriDEX Page 1 of 2 AMEllDMEtiT APPLICATI0tl l10.
78-4 IiiDEX DATE:
4/30/79 Section Remove Add Reason Page Revision Date Page Revision Date III 75 0
4/30/75 75 2
4/30/79 Stacked Storage Added 76 0
9/19/75 76 1
10/13/78 7.4.4.14 Relocated IV App.1 App. 1 Page 3 0
4/30/75 Page 3 2
4/30/79 Table Updated App. 1 App. 1 Page 4 0
9/19/75 Page 4 2
4/30/79 Table Updated V
Index 1 0 4/30/75 Index 1 1
10/13/78 UO Deleted 2
15 0
9/19/75 15 1
10/13/78 Emergency Training 26 0
4/30/75 26 1
10/13/78 QA Audit Requirement Added 28
- 28 29 0
4/30/75 29 1
10/13/78 Redundant Table nso1 Mated Deleted 30 30 s 41 0
4/30/75 41 1
10/13/78 Delete Redundant Isolation Criteria 2
0 9/19/75 1
10/13/78 UO Deleted 2
43 43 44 44 thru 0
4/30/75 thru 1
10/13/78 UO Deleted 2
49 49 50 0
12/01/75 50 1
10/13/78 UO Deleted y
51 1
5/02/77 51 2
10/13/78 UO Deleted 2
52 0
12/01/75 52 1
10/13/7S
- O Deleted g
0 4/30/75 1
10/13/73 UO, Deleted 54 )
54 )
~
55 55 56 0
9/19/75 56 1
10/13/7S UO Deleted 2
57 57 2318 154
Stim-1168 DOCKET 70-1201 REVISIOt1 It1DEX Page 2 of 2 AMEtlDMEtiT APPLICATI0:1 fl0.
78-4 Ii4DEX DATE:
4/30/79 Section Remove Add Reason Page Revision Date Page Revision Date V
58 2
1/24/78 58 3
10/13/73 UO Deleted 2
59 0
12/01/75 59 2
4/30/79 UO Deleted 2
60 -
60 0
4/30/75 1
10/13/78 UO., Deleted -
61 61 J 7.4.4.12B Revised 62 0
9/19/75 62 2
4/30/79 U0, Deleted, Stacking Adaed 78 0
4/30/75 78 1
10/13/73 Contamination -Levels 92 )f 1
0 4/30/75 1
10/13/78 UO Deleted, QA Items 2
92 )
From Deleted Page 28 through 30 Added 94 1
1/24/78 94 3
4/30/79 Contamination Levels 95 1
1/24/78 95 2
10/13/78 Contamination Levels 95a 0
1/24/78 95a 1
10/13/78 Contamination Levels; 8.5 Deleted as Re-dundant. Contained in 19.3 101 0
4/30/75 101 1
10/13/78 8.8.4 Deleted -
Redundant 2318 155
BABC00( & WIL(DX COPAW, C&ERCIAL NUQIAR REL PLAIT USNRC LICENSE SNM-ll68 DOCKET 70-l?01 SECTIG4 III NUCLEAR SAFETY ANALYSIS - PELLETIZING
- 7. '4 Pelletizing 7.4.2 Individual Unit Safety 7.4.2.13 Scrao and Fines Storage (continued)
< 2 weight percent. The maintenance of this internal
_moderator level is assured since the fines are generated in a moderation controlled unit (<l.42 w/o moisture equivalent).
The fiberpacks incorporate internal plastic packaging.
KENO calculations utilizing 16 group Hansen Roach cross sections demonstrate the safety of storage of SNM in any form, when contained in geometrically safe containers (fiberpacks or equivalent) stacked up to 4 units high.
Individual stacked accumulations may be stored adjacent to equivalent accumulations in a linear array when separated by a center-to-center spacing of 28 inches or greater.
Calculational assumptions were as follow:
- L-shaped" array located on two adjacent walls of a room, 18 units per wall in contact with wall, 20 cm concrete wall (actual walls metal), 30 cm concreta floor 1 inch water above.
~ ~ "
- Each unit was a cylinder 9 5/8" diameter x 46" high filled with heterogeneous U(4)02.in water at 70/30 volume percent water and U0 respectively.
2 KEN 0 results C-T-C Spacing (inches)
K-effective + 2c 34
.908 +.008 30
.914 T.006 28
.922 T.008 26
.947 T.007 25
.950I.008 7.4.2.14 Movement of Sm!
The movement of large quantities of SNM, under modera-tion control, has been discussed above in connection 2318 156 DATE 4/30/79 REVIS!01 NO.
2 PAGE 75 CURRENT REVISIGi:
- Added.
SUPERSEEES: PAGE 75 USNRC APPROVAL REFERENCE 4 30/75 ggyg3;g; O
DATE
PRC00< & WILCDX 01 PAW, CEEKIAL NLGEAR REL PIM LENRC LIENSE SNti-ll68 DOCKET 70-LT1 III NUCLEAR SAFETY ANALYSIS - PELLETIZING SECTIQ4 7.4 Pelletizino 7.4.2 Individual Unit Safety 7.4.4.14 Movement of Smit (continued) with the transport containers. Operation of the pelletizing area will require the transfer of smaller accumulaticns of pellets or powder between process steps or into product or scrap storage.
Three basic types of movement configurations are envisioned:
(1) Movement of scrap contained in fiberpacks (or equivalent geometrically safe containers); (2) move-ment of in-process pellets in 4.0 inch slabs, and (3) movement of " completed" pellets into storage prior to rod loading.
Nuclear safety parameters for the movement of SNii were developed as a part of the interaction cal-culations (Part 4.3).
The KEN 0 calculations demonstrate that an "in-transit" fiberpack (containing heterogeneous, optimum 1y moderated U0 ) is safe when positioned 7
12 inches (30.5 cm) from, and at the center of, the in-process pellet storage rack. As noted above, the pellet storage rack is the most reactive single unit in the array.
It follows, therefore, that a one foot separation vetween in-transit fiberpacks and any other accumulation satisfies nuclear safety requirements. Transport carts are limited to one fiberpack positioned so that cart dimensions provide a minimum one foot separation from other accumulations.
4.0 inch pellet slabs may be transported in single layers on carts having a total surface area of 6.25 square feet.
In constructing the KENO model,
" loaded" pellet transfer carts were assumed to be in contact with the following units:
2318 157 DATE 10/13/73 REVIS!C1 NO I
PAGE 76 CURRENT REVISICN:
- " Reloca ted.
SUPERSENS: PAGE 76 USNRC APPROVAL REFERENCE DATE 9/I9/75 PCVISIG:
0
BABOXX & WIL(DX CDP /M, GtERCIAL NUGBR REL PLM USNRC LICENSE SNft-ll68 DOCKET 70-1201 SECTIQ1 IV HEALTH PHYSICS APPENDIX 1, Page 3
SUMMARY
OF SELECTED HEALTH PHYSICS CRITERIA B.
EXTERtlAL DOSIMETRY (continued) 4.
Neutron Dosimetry a.
Application
- Personnel utilizing neutron generating devices
- Determined by Health-Safety b.
System
- Film, vendor supplied c.
Frequency
- Monthly change or more fre-quently, if required 5.
Criticality Dosimetry a.
Personnel
- Indium foil incorporated in TLD/or ID badge b.
General
- Criticality dosimeters placed throughout plant
- Contains AU, IN, S, TLD and glass dosimeters
- Analytic capability available on site C.
SURFACE CONTAMINATION CONTROL Items 1, 2, and 3 of this section represent contamination levels which, when exceeded, will cause clean-up activities of the affected area to
,f, be initiated.
1.
Controlled Areas 2
a.
Removable =
7,500 DPM/100 cm b.
Total =
50,000 DPM/50 cm c.
Survey Frequency
- Weekly 2318 158 l
nu pe n u..s i
DATE 4/30/79 l REVISICN f:0.
2 PAGE 3
- "Leveis/ Areas Adusted.
CURREW REVISICd:
SUPERSELES: pac-E APPe" ix I' P39e 3 m mm paa 4/30/75 DATE REVISICT1
B20XX & WIL(DX C&PA'h', CTEfCIAL NUCEAR REL PUM USNRC LICENSE SNM-ll68 DOCKET 70-131 SECTIQ1 IV HEALTH PHYSICS APPENDIX 1, Page 4
SUMMARY
OF SELECTED HEALTH PHYSICS CRITERIA C.
SURFACE CONTAMINATION CONTROL (continued) 2.
Change Rooms (Intermediate Area) 2
- a. Removable =
500 DPM/100 cm 2
b.
Total =
2,000 DPM/ 50 cm c.
Survey Frequency
- Daily when in operation I
3.
General Plant Area a.
Acceptable average 2
removable =
- 200 DPM/100 cm 2
500 DPM/50 cm b.
Total =
c.
Survey Frequency
- Monthly or more frequently as required 4.
Materials and Equipment a.
Transferred from controlled 2
area
- 100 DPM/100 cm removable ***
for unrestricted release
- Higher levels under special controls by Health-Safety b.
Survey Frequency
- As required 5.
k'aste and Scrap for Unrestricted Release 2
a.
Maximum Removable =
1 1000 DPM/100 cm 2
b.
Maximum Total =
1 12,500 DPM/50 cm 2
c.
Average Total =
< 2,500 DPM/50 cm d.
Beta & Gamma 1 1 mrad / hour (Average 0.2 mrad /hr) e.
Survey Frequency
- Monthly or mcre frequently ne cen. ic a a y e. L e i sa 4 A DATE 4/30/79 REVISIGJ f.'O.
2 p;GE 4 CURRENT REVISICN:
SLFERSEms: PAGE AoPendix 1 P39e 4 wNRCAPPQ/(REFERENCE
/ U/ 3 REVIS101 0
d I8 i59 DATE
BABCDO( & WIL(DX COPA'h', C@iERCIAL NUClfAR PE. PLAIT USNRC LICENSE SNM-ll68 DOCFET 70-1201 SECTIO 4 V CONDITIONS Page Part 1
1 Scope 2
2 Authorized Activities 3
3 Authorized Materials 4
4 Glossary of Terms 6
5 Administrative 6
5.1 fianagement 6
5.2 Operational Area Supervision
~
7 5.3 Health-Safety Section 9
5.4 Independent Auditors 10 5.5 Nuclear Criticality Safety Analysis 11 6
General Specifications 11 6.1 Control Program 14 6.2 Emergency Plan 16 6.3 Personnel Training 23 6.4 Postings 23 6.5 Enrichment Control 24 6.6 Procedure Control 26 6.7 Audits and Q. A.
31 7.0 Nuclear Safety - Technical Specifications 31 7.1 General Requirements 40 7.2 Material Specifications 41 7.3 Fuel Pellet Receiving & Temporary Storage 42 7.4 Fuel Pellet Fabrication 63 7.5 Pellet Vault Storage 64 7.6 Fuel Pellet Handling & Storage 65 7.7 Fuel Rod Processing & Storage 67 7.8 Reserved 08 7.9 Fuel Assembly Processing 69 7.10 Fuel Assembly Storage & Packaging DATE 10/13/78 ggyIstag go, 1
PAGE 1
CURRENT REVISIG4:
SUPERSEEEs: PAGE re w,
~E~~
DATE 4/30/75 psyisig; O
}
BABC00( & WILCDX C&PRf/, COREPLIAL NUOf# REL PLM LENRC LIENSE Snit-ll68 DOCKET 70-1201 V C0flDITIONS SECTIO 4 6.0 General Specifications 6.2 EMERGENCY PLAT! (continued) d.
Post-accident recovery and reentry actions includi19 guide-lines for implementing these actions including corrective actions that may be necessary to terminate or minimize the consequences of the accident, criteria for plant reentry, physical security, and resumption cf operations.
e.
Arrangements for transportation to and treatment of in-dividuals at facilities outside the site boundarys f.
Provisions for emergency first aid, medical attenti~on, decontamination, and exposure estimation including a definition of available equipment and facilities.
g.
Provision for annual evacuation drills.
h.
Annual familiarization of responsible off-site fire fighting units wi.th plant layout, activities, and nuclear safety controls.
In addition, training for the primary off-site medical support facility will be conducted every two years.
i.
Provision for maintenance and storage of emergency equipment, considering the various types of accidents thac can be anticipated.
Health-Safety is responsible for maintenance and control of the emergency plan.
The plan will be distributed, as appropriate, to plant personnel and other individuals with response assignments.
2318 161 DATE 10/13/78 REVISICtl NO, 1
PAGE 15 CURRENT REVISIC'!:
- 6.2g revised, 6.2h added, 6.2i renumbered.
Sl;FER3 EMS: PAGE I5 USNRC APPROVAL REFERENE 9/19/75 0
DATE REVISIG1
BABC00( & WILCOX CUP /M, C&iOCIAL NUCIBR REL PLM WNRC L! CENSE SNM-ll68 DOCKER 70-]201 SECTica V CONDITIONS 6.0 General Specifications 6.6 Procedure Control (continued)
Revised procedures shall be subject to approval in the same manner as new procedures. Health-Safety procedures will be reviewed at least annually for technical correctness and applicability.
Procedure distribution and control is the responsibility of plant supervision.
6.7 Audits and Quality Assurance An internal audit and quality assurance program shall be maintained to provide assurance that plant activities are conducted safely and in accord with license specifications.
The audit program shall be the responsibility of the Manager, Safety, Licensing, and Safeguards.
Health-Safety supervision shall conduct, at least weekly, a fomal audit of plant status relative to nuclear and radio-logical safety. At the discretion of Health-Safety, the audit may consist of an in depth evaluation of a specific area or may be of a generalized nature providing an overview of total plant activities. Audit results shall be documented, reported to plant management and supervision as appropriate, and will be maintained on file by Health-Safety for at least 6 months.
Health-Safety audits shall be conducted by personnel technically qualified in operational nuclear and radiological safety and in the application of license specifications.
2318 162 DATE 10/13/7S REVISIQJ No, 1
FAGE 26 CURRENT REVISIGJ: "* Relocated frca Pace 23 SUPERSEEES: PAGE 26 WNRC APPRCVAL REFERENCE DATE J/30/75 psyysigg 0
I PRCDO( a WILCOX O1P/M, C&ERCIAL NUCLEAR REL PUM LEtJRC LI tJSE StCt-ll68 DOCKET DJ201 V C0flDITI0tl5 SECTICt1 PAGES 28, 29, and 30 It1TEt1TI0tlALLY LEFT BLAT 1K.
2318 163 I
DATE 10/13/70 REVISICt; t;3, PACE s 28, 29, & 20
- Deleted - Recundant..
CURREt1T REVIS!Cf t:
SUPERSErcs: PAGE S 28, 29, & 30 USNRC APPROVAL FEFEREI;G g
DATE 4/30/75 REVISICt1
BMGCK 8 WILGX C&PM, CCitERCIAL NIHEAR REL PUM tENRC LICENSE Snit-ll68 DOCKET /D l%_ 1 SECTlag V C0t1DITI0t15 7.0 t1uclear Safety - Technical Saecifications 7.3 Fuel Pellet Receivino and Temocrary Storace A single shipment of pellets may be stored in the shipping containers in an array no more reactive than as received pending transfer to planned storage facilities. The shipment will be isolated from other SilM as specified in 7.1.2.
7.3.1 The containers will be examined for transit damage, and Health-Safety notified if damage is found.
Heal th-Safety will examine the damaged containers for evidence of water entrainment and other cond'tions that might constitute a hazard. Damaged containers will be opened one at a time and as the material is removed, it will be immediately placed into storage within safe geometry slabs or 850 grams U235 accumulations in accord with appropriate license conditions for storage.
7.3.2 Damaged containers containing Stim will not be recurned to the shipper.
2318 164 I
CATE 10/13/73 REVISICr4 tio, PAGE CURRENT REVISICN:
SUPERSEEES: PAGE al USNRC AFFRCVAL REFERENCE
/ 0/,m 0
DATE REVISICtf
B/BC00( a WILQX 01PRf(, CCRERCI/L NUCLEAR REL PLM USNRC LICENSE SNM-ll68 DOCKET 70-LM1 V CONDITIONS SECT!al
'7.0 Nuclear Safety - Technical Soecifications 7.4 Fuel Pellet Fabrication This part describes the license conditions specifically applicable to the fuel pellet manufacture at the CNFP.
7.4.1 Changes and modifications undertaken in the pellet fabrication area shall be conducted only in accord with the procedural specifications and limitations set forth in Part 7.1.4.3 of this section.
7.4.2 In addition to the definitions presented in Part 4.0, the following definitions apply mecifically to the pellet fabrication area.
7.4.2.1 Calculated Saf: Units - are those units evaluated according to Part 7.1.4.3 and shown to have a Keff not exceeding 0.85 under normal operating conditions and 0.95 under assumed accident conditions, con-sidering both statistical and methodology limit of error.
7.4.2.2 Calculated Safe Array - is one evaluated according to Part 7.1.4.3 and shown to have a Keff not ex-ceeding 0.90 under normal operating conditions 2318 105 CATE 10/13/13 REVISICN NO.
1 PAGE 42 CURRENT REVISICft:
SUPERSEES: PAGE J2 LENRC AFFROVAL FEFERENCE DATE C/10'75 REVIS!01
B20XX & WILCOX OIPM, CatEPflAL NUClfAR FLEL PIM USt4RC LICEf4SE Stat-ll68 DOCKET 70-1201 SECTIGa V
C0t1D'TI0 tis 7.0 fluclear Safety - Technical Scecifications 7.4 Fuel Pellet Fabrication 7.4.4.2 and 0.95 under assumed accident conditions, considering both statistical and methodology limit of error.
7.4.3 General Moderation Control Specifications This part provides the general criteria to be applied for assuring the maintenance of acceptable H/U levels within moderation controlled equipment, and the limitation of interspersed moderation within the pelletizing area as a whole.
In some cases, additional moderation control conditions are imposed for specific units and are incorporated in the conditions for that unit.
2318 166 CATE 10/13/78 REVISIQit;0, PAGE 33 CURFEtiT REVISICf!:
SUPERSEDES: PAGE a3 u3;;RC APPROVAL REFEREliCE DATE 9/19/75 REVISICt1 0
BABC0(X & WILQX CRPRt/, CMERCIAL NUClfAR FLEL PUitT LENRC LICENSE SNtt-ll68 DOCKa. 70-1201 SECTIOi V CONDITIONS 7.0 Nuclear Safety - Technical Soecifications 7.4 Fuel Pellet Fabrication 7.4.3.1 Area tioderation Control The following conditions are specified to limit the potential for the generation of significant levels of interspersed area moderation and to provide protection a gainst the accidental introduction of hydrogeneous materials into moderation controlled units and arrays.
A.
Service water piping entering the pel-letizing area shall be baffled or enclosed in non-flammable conduit so as to preclude spraying in the event of a joint or line rupture.
B.
Those process steps requiring the use of liquid moderators (e.g., furnaces,
coolants, and grinding liquids) shall incorporate integral or adjacent, non-flammable shielding, baffling, or equivalent measures to preclude general area moderation in the event of a failure.
C.
flo water lines or other potential liquid moderator sources shall be located directly above moderation controlled units in a position such that a failure could result in internal moderation of DATE 10/13/73 REVISICN ta, 1
PAGE 44 CURRENT REVISION:
SUPERSE:ES: PAGE 44 USNRC APPRCVAL FEFERENCElg lg7 DATE 4/30/75 REV!SION 0
BfBC00( 8 WILOM 01P/M, CCftEPflAL NUCLEAR REL PUM USNPC L! CENSE SNti-ll68 DOCKET 70-1. l T SECTrat V C0flDITI0 tis 7.0 fluclear Safety - Technical Scecifications 7.4 Fuel Pellet Fabrication 7.4.3.1 Area floderation Control (continued) C. Continued that unit. D. Monthly inspections shall be conducted by Health-Safety to verify the continued integrity of baffles, seals, shields, or other protective devices. E. The contents of any vessel other than process units containing liquid modera-tors shall be limited to 3.5 gallons. F. A fire protection program will be de-veloped such that - Fire fighting agents available in the area will be limited to non-hydrogeneous types such as CO, "Halon," or dry chemical. 2 Water sprinkler systems are pro-hibited. - Flammable wastes are accumulated in fireproof containers and are removed from the area daily. - The plant fire brigade and area fire departments will be famil-iarized with the restrictions on the use of hydrogeneous agents 2318 108 DATE 10/13/73 REVISIQ1 NJ. 1 PAGE 45 CURRENT REVISICU: SUPERSELES: PAGE 45 USNRC APPROVAL REFERENCE DATE 4/30/75 REVISICtj 0
BABC00( & WIL(DX CDP 4lY, C&EKIAL NUCLEAR FLEL PLUT USNRC LICENSE SNM-ll68 DOCKET 70-1201 SECTlai V C0tlDITI0tlS 7.0 fluclear Safety - Technical Specifications 7.4 Fuel Pellet Fabrication 7.4.3.1 Area Moderation Control (continued) F. Continued in the area and prominent postings displayed specifying the applicable controls. 7.4.3.2 Unit tioderation Control The following criteria establish specific requirements that must be satisfied for all powder operations under moderation control. As appropr.: ate, additional requirements are noteu in 7.4.4 for individual units. A. The average weight percent water equi-valent in the blender, transport con-tainers, and powder preparation unit shall not exceed 1.42 (H/U = 0.45). B. 'a weight percent moderator content of ii. coming powder shall be determined, as socified in 7.4.3.3, prior to loading into the blender. The moisture content of recycled powder will also be detemined, on a "by container" basis, prior to loading into a moderation con-trolled unit if the material has been 2318 109
== DATE 10/13/73 REVISIOJ NO. 1 PACE 46 m CURRENT REVISICN: SUPERSEES: PAGE 46 LENRC AFFROVAL REFERE! ICE DATE 4/30/75 REVISICri
BA3C00( & WILQX CaPM/, C&iERCIAL NUCifAR REL PLM USNRC LICENSE SNM-ll68 DOC!ET70-lN_1 SECTIa1 V CONDITIONS 7.0 Nuclear Safety - Technical Specifications 7.4 Fuel Pellet Fabrication 7.4.3.2 Unit Moderation Control (continued) B. Continued exposed to non-moderation controlled conditions. C. Positive mechanical connections or enclosures will be utilized for the transfer of oxide between moderation controlled units or process steps. D. Other than when required for opera-tion, all penetrations into modera-tion controlled units shall be closed. Infrequently used penetrations shall be capped, or bolted (or screw type) flanges installed. E. Whenever practicable, the decontamina-tion of moderation controlled vessel interiors shall be accomplished using dry techniques. If the use of liquid agents is required, the unit shall be inspected by Health-Safety and a written certification prepared that visible moisture is not present. The accumula-tion of decontamination liquids in any single vessel in the vicinity of modera-tien controlled units shall be limited to 3.5 gallons (safe volume). 2318 170 ]'13/73 REVISICJ No, I PAC 3_ 47 DATE CURRE:!T REVISICf1: SUFERSEEES: PACE 47 USNRC AFFROVAL REFERENCE 0 DATE 4/30/75 REVISICt1
BABC00( & WIL(DX CUPRU, CUiEFCIAL NUCLEAR REL PLAIT USNRC LICENSE SNtt-ll68 DOCKET 70-lT_1 SECTIQi V CONDITIONS 7.0 Nuclear Safety - Technical Specifications 7.4 Fuel Pellet Fabrication 7.4.3.3 Moderation Content Verification and Sampling 0xide received from the vendor is sampled and analyzed after receipt at the CNFP. A. For nuclear safety purposes, the follow-ing procedure will be followed in deter-mining the moisture content of incoming powder: 1. A " thief" (or equivalent) sample will be collected from each powder container upon receipt at the CNFP. 2. A composite sample representative of not more than 30 containers will be analyzed to detennine the as-received internal moderator level. All containers in a shipment shall be sampled and analyzed in accord with the above program. Minimum detection limit shall be 0.1 weight percent water equivalent. 3. Analytical results must confirm that the average weight percent water equivalent in any single powder lot does not exceed 1.42 prior to 2318 171 DATE 10/13/78 REVISIQ1 NO. 1 PACE dS CUR.EiT REVISIGJ: SUPERSEEES: PAGE 43 USNRC APPRCVAL REFERENCE DATE 4/30/75 pgyy3;g; O
BABC00( & WILGX COPAW, QMERCIAL NUGIAR REL PUWT USt4RC LICENSE SNti-il68 IXXXEr 70-1201 SECTICr4 V r0NDITIONS 7.0 Nuclear Safety - Technical Soecifications 7.4 Fuel Pellet Fabrication 7.4.3.3 Moderation Content Verification and Sampling (continued) A. 3. Continued loading into a moderation controlled unit. Individual containers with a moisture content of 1 2.0 weight per-cent may be added to the blender lot if the 1.42 w/o average is not exceeded. 4. Powder lots or individual containers not satisfying moisture content spe-cifications are identified and con-trolled to preclude inadvertent ad-dition to a controlled unit. B. Powder Lubricant Control The addition of powder lubricant to powder in the blender or transport container is discussed in 7.4.4.2 and 7.4.4.3 below. The following specific conditions apply: 2318 172 DATE 10/13/78 REVISIG4 NO. I PAC {
- 9 CURRENT REVISICil:
SUPER 3EM3: PAGE 49 LSt!RC AFFROVAL REFERENCE DATE 4/30/M pgygggg; O
B/ ECOG ( & WILQX CUPA'f/, CUiERCIAL NIEEAR FLEL PLM usNRC uCENSE SNft-ll68 IDCKET 70-1201 SECTIQ; V CONDITIONS 7.0 Nuclear Safety _- Technical Soecifications 7.4 Fuel Pellet Fabrication 7.4.4.3 Moderatio-Content Verification and Sampling (continued) B. Powder Lubricant Control (cont'd) 1. A sample of the blender or transport container contents will be analyzed and the t0tal weight percent moisture content, including lubricant con-tribution, calculated prior to any lubricant addition. 2. Addition of powder lubricants to the process stream is authorized only upon verification that the H/V in the receiving unit will not exceed 0.45. 3. For calculational purposes, the hydro-gen content of powder lubricant will be assumed to be 1.2 times that of water. 4. Lubricants will be pre-packed in 3 kg or smaller aliquots before issue. C. Administrative Control 1. Moderator content evaluations shall be overchecked for completeness and accuracy by an individual other than the one perfoming the original analysis prior to further processing of the material. }}]g }[} DATE 10/12/7s REyISla; NO, 1 FAGE 50 CURRENT REVIS!Cft: SLFERSEEES: PAGE 50 USNRC AFFROVAL REFERENCE O DATE 12/1/75 pgyI31g;
BtEQ0( & WILCOX C[tP/M, CUtERCIAL NUCLEAR REL PLM WNRC LICENSE SNtt-llo8 DOCKET h]201 V CONDITIONS SECTIO 4 7.0 fluclear Safety - Technical Scec_ifi_c_ations 7.4 Fuel Pellet Fabrication 7.4.4.3 Moderation Content Verification and Sampling (Continued) C. Administrative Control (cont'd) 2. Monthly audits shall be conducted by Health-Safety to verify con-tinued proper functioning of modera-tion control systems and procedures. 7.4.4 Individual _U_ nit Safety The conditions in this part define the specific nuclear safety criteria to be applied to individual operations and process units within the pelletizing area. Slab thickness specified below may not be increased with-out license amendment. The arrangement and size of accumulations within the area may be varied except as limited in the specific conditions outlined below, and provided the requirements of 7.4.1 above are satisfied. 7.4.4.1 Sf!!! Receipt and Storage A. SNM may be temporarily stored in authorized shipping containers in an array that is no more reactive than that specified for the same material under transport conditions. In ad-dition, individual containers (fiberpacks or ecuivalent) may be stored in the DATE 10/13/73 REVISIa1 No, 2 PAGE SI CURRENT REVISIG!: SUPERSErEs: PACE 51 USNRC AFPROVAL REFERE? ICE 5/2/77 pgy;3gg; 1 DATE 2318 174
BABC00( & WILOM CUPA'h', CSERCIAL NUClIAR REL PLM USNRC LICENSE SNi4-ll68 DOCKET 70.1201 SECTIG4 V CONDITIONS 7.0 Nuclear Safety - Technical Soecifications 7.4 Fuel Pellet Fabrication 7.4.4.1 SNf1 Receipt and Storace A. Continued receiving area when separated from adjacent units by a minimum of 30 inches center-to-center. B. Powder in fiberpacks (or equivalent) will be transferred to the powder sampling station in a single horizontal row. The powder sam-pling station array shall be limited to 7 upright containers. C. - Damaged containers that may have the inte-grity of the inner package compromised and could be responsible for increased modera-tion shall be segregated on the receiving dock and spaced on 30-inch centers in a single-plane array. - Damaged containers, as described, may be brought into the sampling hood for repack-aging and/or sampling while there are no other SNM-bearing containers in the samp-ling and powder storage-rack loading areas. Only one such container may be in this area at any one time. - The damaged containers, as described, will not be placed on the powcer storage rack unless the moisture content u no greater than 2 weight percent. D. In-plant storage of powder in sealed con-tainers shall be such that the containers, laying on their sides, form a 2 wide by 5 high by 60 foot long hori ontal array. A minimum 22 inch center-to-center spacing will be maintained between adjacent rows of containers. - Arrangement of powder containers on the hoists serving the storage rack shall be the same as on the storage rack, or, where the powder weight cercent moisture content satisfies the specification of 7.4.3.2 A of this section, may be limited to a single row of 5 upright containers. DATE 10/13/73 REV!SI W NO. I PACE 52 CURRE:,7 REVISICN: SUPERSEDES: PAGE 52 USNRC APPROVAL REFEFEJCE DATE 12/1/75 REVISICU 0 2318 175
BABC00( 8 WILQX C& PAW, C&iERCIAL NUClfAR REL PLM IENRC LI&NSE SNM-ll68 DOCKET 70-lN_1 V CONDITIONS SECTIG1 7.0 Nuclear Safety - Technical Soecifications 7.4 Fuel Pellet Fabrication 7.4.4.2 Powder Blending A. The powder blender is limited to 3600 kg of oxide. The weight percent water equivalent of the powder shall be de-tennined prior to loading into the blender and shall not exceed that specified in Section V, Part 7.4.3.2. B. The blender feed port shall be enclosed by a non-flammable glovebox. C. Non-hydrogeneous lubricants shall be used for internal portions of the blender mechanism. D. Powder lubricants may be added to the blender contents provided the following criteria are satisfied: - The blender has been operated to assure homogeneity, and the con-tents sampled and analyzed to determine the average weight per-cent moisture present. - A verifying calculation perfonned to assure that the ambient moisture content plus the proposed lubricant contribution will not exceed 1.42 w/o. 2318 176 DATE 10/13/75 REVISIGJ NC. 1 PAGE 53 CURRENT REVISICf!: SLFERSENS: PAGE 53 USNRC APPROVAL REFERENCE DATE 4/30/75 REVIS!Q; O
BEC00( & WIL(DX CUPAh', CatEEIAL NUCLEAR REL PIM USNRC LICENSE SN?t-ll68 IDCKET 70.1201 SECTIO 4 V CONDITIONS 7.0 Nuclear Safety - Technical Specifications 7.4 Fuel Pellet Fabrication 7.4.4.2 Powder Blending (continued) D. - Powder lubricant shall be added in increments not exceeding 1 kg per minute and with the blender operating. 7.4.4.3 Powder Transport (Transport Container) The powder transport containers shall be operated under the same internal moderation controls applied to the blender, and may be utilized for the following activities: - Powder receiver and feed vessels for the blender (receiving only), powder preparation, and pelletizing (feed only). In addition, the transport containers provide in-process storage capacity and may serve as supplemental lubricant addition and blending units. A. The maximum internal volume available for powder shall be 600 liters. B. Use of transport containers will be per-mitted only in relation to the blender, powder prepara tion units, tumblers, and pellet press. The total number of trans-port containers is limited to 7. C. Powder lubricants may be added to 2318 177 DATE 10/13/78 REVIS!Q: NO. I PAGE 54 CURRENT FEVISICf1: SUPERSEreS: PAGE 54 LGNRC APPROVAL REFERENCE DATE 4/30/75 REVISIQ4 0
BABC00< & WILWX CUP ##, C@iERCIAL NUCEAR FLEL PLAIT USNRC LICENSE SNft-ll68 DOCKET 70-1201 SECTICt1 V CONDITIONS 7.0 Nuclear Safety - Technical Soecifications 7.4 Fuel Pellet Fabrication 7.4.4.3 Powder Transport (Transport Containers) (cont'd) C. Continued transport containers when the mass of the contained oxide does not exceed 1000 kg. Lubricants will be added in maximum aliquots of 3 kg following sampling to verify that the total weight percent moisture equiva-lent will not exceed that specified in 7.4.3.2. 7.4.4.4 Powder Preparation (Slug Pressing, Granulating, Powder Size Characterization) A. Powder accumulation within the powder pre-paration units shall not exceed 1300 kg (except in associated transport containers). Internal moderation of powder shall be limited to that specified in 7.4.3.2. B. Powder preparation equipment shall be hooded or otherwise enclosed to preclude external moderation. 2318 178 DATE 10/13/7? REVISICtl NO. 1 PAGE 55 CURRENT REVISICt': SUPERSErcs: PAcg 55 USNRC APPROVAL FEFERENCE DATE 9/19/75 REVISICtl 0
BABWO( 8 WILWX C&P#t/, C0FTERCIAL NIEEAR REL PIM USNRC LIENSE Stitt-ll68 EOCKET 70-1201 SECTIQ4 V C0tlDITIONS 7.0 tiuclear Safety - Technical Soecifications 7.4 Fuel Pellet Fabrication 7.4.4.5 Powder Pressing The pellet press and associated pellet handling surfaces shall not exceed a maximum slab thickness of 4 inches and shall be limited to an area of 54 square feet. 7.4.4.6 Sintering A. Sintering furnaces shall be limited to a 6 inch high slab of pellets. Maximum slab width shall be 12 inches. B. Pellet handling supporting the sintering operation shall be limited to a total available surface area of 22 square feet per furnace. Maximum slab height shall be 6 inches. 2318 179 DATE 10/13/73 REVISICN tio. I PAGE 56 CURRENT REV1SICf': SUPERSEmS: PAGE UStiRC APPROVAL RE E RE!JG 9/19/75 0 DATE REVISIG1
BABC00( & WILOX C& PAW, C0ffEPflAL NUCLAR REL PLM USNRC LICENSE SNft-ll68 DOCKER 70-1201 SECTIQ4 V CONDITI0 tis 7.0 Nuclear Safety - Technical Specifications 7.4 Fuel Pellet Fabrication 7.4.4.7 In-Process Pellet Storage In-process pellets shall be stored on racks so that each rack is limited to four 4.0 inch thick slabs 26 inches wide by 24 feet long. Shelves are separated vertically by 16 inches center-to-center. No more than two such racks shall be installed in the area. Shielding shall be installed above the racks to pre-clude moderation from potential sources located directly above the rack. 7.4.4.8 Pellet Grinding The pellet grinding operation is taken to include collateral activities such as grinder sludge and coolant control. The activities will be such that: A. A maximum 4 inch slab will be authorized for accumulation of pellets. Slab area, excluding grinder and its associated feed and discharge mechanisms, shall not exceed 18 square feet. 2318 180 DATE 10/13/7S REVISIci No, 1 PAGE 57 CURF5'IT REVISICN: *'* SUPERSEEES: PAGE 57 WNRC APPROVAL REFEFSICE DATE 9/19/75 REVISIG4 0
BABC00( & WILQX CCf PM/, C0fiERCIAL NUCLEAR REL PIM LENRC LICENSE SNM-ll68 DOCKET 70-J201 SECTIG4 V CONDITI0tl5 7.0 Nuclear Safety - Technical Specifications 7.4 Fuel Pellet Fabricatic.1 7.4.4.8 Pellet Grinding (continued) B. Units controlled to a safe volume criteria will be limited to a maximum capacity of la liters (e.g., centrifuge bowls and coolant / reservoir tanks). C. Arrangement of individual units within the grinder complex shall be based on solid angle criteria. D. SNM in the grinder sludge drying oven shall be limited to three full centri-fuge bowls (or equivalent) in a single layer. 7.4.4.9 Dry Scrap Conversion A. The dry scrap conversion furnace shall be limited to an 8.7 inch inside diameter. B. Feed and discharge vessels shall be limited to 9-5/8 inch inside diameter. C. System safety shall be based on solid angle criteria. 2318 181 DATE 10/13/73 REVISICt1 NO. 3 PAGE 58 CURRENT REVISICft: SUPERSEES : PAGE SS LEiRC AoFROVAL FEFERENCE DATE l-24-77 REVISIG4 2
BEC00( & WILCOX COPAW, CMERCIAL NUCLAR FLEL PLM US?4RC LICENSE SNtt-ll68 DOCKET 70-1201 SECTlai V C0tIDITIONS 7.0 Nuclear Safety - Technical Soecifications 7.4 Fuel Pellet Fabrication 7.4.4.10 Packaging Hood A. Limited to two 25 kg or smaller accumu-lations in any fona and arrange-ment. 12 inch edge-to-edge separation shall be maintained between the accumu-lations. 7.4.4.11 Process Control Area and Quality Control Area A. Limited to three 25 kg accumulations of oxide per operating area, in any form and any arrangement, but with a minimum of 12 inch spacing, edge-to-edge, between accumulations, or B. One 4 inch slab, not exceeding 6.25 square feet, per operating area, C. Mass and geometric controls shall not be applied simultaneously within an opera-ting area. 7.4.4.12 SNM Movement Movement of small accumulations af SNM within the pelletizing area shall be limited as follows: A. Accumulations of in-process pellets shall be transported in 4.0 inch slab limited so that the transport array does not occupy an area greater than 6.25 square feet. 2318 182 DATE 4/30/79 2 ggy;3;g; p;g, PAT-59 CLERE:,7 REVISIG1: SUFERSEDES: PAGE WNRC APPROVAL EFEREi1CE I /I/7 DATE REVIS101
BABC00( & WILCDX C&PA#, Ca?EPflAL ECI. EAR FLEL PLM USNRC LIENSE SNft-3168 DCCKET 70-1201 SECTIQ4 V C0f1DITI0f15 7.0 Nuclear Safety - Technical Scecifications 7.4 Fuel Pellet Fabrication 7.4.4.12 SNM Movement (continued) B. Geometrically safe containers (fiberpacks, centrifuge bowls or equivalent) of powder, pellets or scrap in any form may be transported under controls that assure a minimum of 12 inch edge-to-edge spacing between the container and other accumula-tions of SNM. Stacking on transport. carts is not authorized. C. The pelletizing area shall be limited to no more than 6 transport carts. 7.4.4.13 Surge Pellet Storage Shelves and Pelle.t Eox Conveyor A. The surge pellet storage shelves shall be limited to a single vertical array consisting of five 18 inch wide shelves for storage of safe geometry slabs. Total shelf length shall not exceed 36 feet. Vertical spacing between shelves will be 16 inches. Nuclear poison equi-valent to 35 wt. % B C in Al 0.168 inch 4 thick, and having a minimum density of 2.46 g/cc, shall be installed on each shelf except the bottom one. Only a single slab thickness will be permitted on any one shelf. 2318 183 60 CATE 10/13/78 ggyIgga; go, 1 pAgg CURREilT REVISIG1: SUPERSEEs: pac-E 60 USNRC AFPROVAL FEFERENCE DATE 4/30/75 0 ggygggg;
BABOXX & WILWX COPA7(, C&iEEIAL NUCLEAR REL PLM USNRC LICENSE SNM-ll68 DOCKET 70-1201 'l CONDITIONS SECTIG4 7.'0 Nuclear Safety - Technical Specifications 7.4 Fuel Pellet Fabrication 7.4.4.13 Surge Pellet Storage Shelves and Pellet Box Conveyor (continued) B. A pellet box conveyor may be substituted for one of the shelves providing all other criteria are satisfied. The conveyor may extend the length of the pellet fabrication area. Portions of the conveyor extending beyond the storage shelves are not poisoned. 7.4.4.14 Fines and Scrap Collection and Storage A. Geometrically safe containers for the accumulation of process scrap may be positioned as required for operation at individual process units. A mini-mum 24 inch center-to-center spacing will be maintained between containers. B. Accumulations of scrap in geometrically safe containers may be stored at desig-nated locations within the pelletizing area provided the following require-ments are satisfied: - Individual accumulations consist of no more than 8 containers on 24 inch center-to-center spacing and in a 2 wide x 4 long array. 2318 184 DATE 10/13/73 I PAGE 61 REVIS!Cri NJ. CURRENT REVISICN: SUPERSEEES: PAGE 61 usNRC AFrRovAL REFERENCE 4/30/75 0 DATE REVISICt1
BABC00( a WILOM CUP /M, CafERCIAL NUQfAR REL PUNT USNRC LICEf1SE SNM-ll68 DOCVET 70-1201 SECTlai V CONDITI0tlS 7.0 fluclear Safety - Technical Soecifications 7.4 Fuel Pellet Fabrication 7.4.4.14 Fines and Scrap Collection and Storage - Array geometries differing from the above may be used if confirmed by equivalent solid angle calculations. C. The location of storage arrays shall be in accord with the interaction bases established in Section III. D. Scrap storage accumulations positioned between the powder storage rack and the powder preparation operation shall be limited to material generated within moderation controlled units. All other scrap storage may consist of SNf1 in any fonn. E. Vacuum cleaners used for cleanup /decon-tamination purposes in the pelletizing area shall satisfy the safe geometric limits specified in 4.9 or 4.10 of this section. F. Individual geometrically safe containers (fiberpacks or equivalent) containing SNM in any form may be stored stacked up to 4 containers in height in an "L shaped array con-sisting of a maximum of 13 stacks in each leg, with a minicua center-to-center spacing of 23 inches between individual stacked accumulations within the array. Nuclear isolation will be main-tained between the storage array and adjacent S."':-bearing units 2318 185 EATE REVISIGJ NO. 2 PAGE 6? CURREtiT REVISIG1:
- 9dded.
SUPERSEDEC' PAGE 62 USNRC APPRCVAL REFEREi!CE DATE 9/19/75 ggyj 3 g g, O
BABC00( 8 WILODX C& PAW, C&ERCIAL NUClfR FLEL PLAff USNRC LICENSE SNtell68 DOCVEr 70-1201 V CONDITIONS SECTIU4 8.0 Technical Specifications - Radiation Safety 8.1 Effluents to Unrestricted Areas 8.1.3 The following parameters are established for detennining if waste or excess materials and equipment may be disposed of in routine industrial fashion. If any of the specifications set forth below are exceeded, the material will be controlled as contaminated. a. The maximum total alpha contamination will not
- t*
exceed 12,500 disintegrations per minute per 50 square centimeters. Total contamination shall be defined as fixed + removable. b. The average total alpha contamination on any item of waste or equipment will not exceed 2500 disintegrations per minute per 50 square centi-l meters. c. The maximum alpha contamination removable by wipe will not exceed 1000 disintegrations per minute per 100 square centimeters. d. Beta + gamma radiation level when measured with an open shield, thin-tube wall G-M survey instru-ment at one centimeter will not exceed a maximum of one millirad per hour, nor an average of 0.2 mrad /hr. 2318 186 DATE 10/13/73 REVISICri No, I PAGE 73 "* Rewo rd ed. CURREP.T FEVISIC(1: SUPER $Em S: PAC-E 74 USNRC APPRCVAL REFERENCE 0 DATE 4/30/75 REVISIG1
BABC00( & WILOX C& PAW, CUTERCIAL NUQfAR FlEL PlM USf4RC LICEf4SE Stitt-]l68 DoCxET 70-]201 SECTIO 4 V C0fiDITI0ftS 8.0 Technical Specifications - Radiation Safety 8.3 Instrumentation (continued) 8.3.4 All instrumentation will be calibrated upon initial installation, following major maintenance, and at other times as' deemed necessary. In any case, calibration will be perfonned at least semi-annually. 8.3.5 Field check sources will be available for assuring functional response of instrumentation prior to use. 8.4 Personnel Monitorino and Contamination Control 8.4.1 Personnel monitoring equipment - film badges, thenno-luminescent dosimeters, detectors, etc., - will be used by persons having access to restricted or radi-ation areas, as deemed necessary by Health-Safety, and in accord with 10 CFR 20. The dosimeters will be processed routinely to determine radiation dose. Health-Safety will review and maintain the dose records and prepare such reports as are required by regulations. fleutron dosimeters will be used by operators and other persons as deemed necessary by Health-Safety when a potential for measurable neutron exposure exists. 2318 187 DATE 10/13/7S REVISICtl t:0, I PAGE 91 CL'RREt.T REVISICf1: SLFERSEDES: PAes 91 USNRC APPROVAL FSFEFBiCE DATE /30// 0 REVISIcti
BABC00( 8 WIL(DX C&PN#, OMERCIAL NUQfAR REL PUilT USNRC LICENSE SNM-ll68 DOCKET 70-1201 SECTICt1 V CONDITIONS 8.0 Technical Specifications - Radiation Safety 8.4 Personnel Monitoring and Contamination Control (cont'd) 8.4.2 An indium foil strip will be worn by plant personnel to provide for identification of persons receiving high exposures in the event of an accidental criticality. 8.4.3 Persons who work routinely in areas where there is a potential for bodily intake of radioactive materials will be subject to determination of the extent of in-take and retention by analysis of excreta or body counting. The bicassay program shall be conducted in accord with Regulatory Guide 8.11, " Application of Bioassay for Uranium". Periodic " blank" or " spike" samples will be included in the bicassay program. 8.4.4 Personnel Contamination " Change room" facilities shall be provided for the control of contamination spread from areas where sig-nificant quantities of unclad SNH are handled. Nomal exit from controlled areas shall be through change room facilities. The change room shall include personnel decontamination and monitoring capability. External contamination of personnel entering controlled areas shall be limited by means of protective clothing. The degree and type of protection required is determined by Health-Safety, depending on the contamination potential 2318 188 DATE 10/13/73 REVISICU !:0. 1 PAC { 92 CURFE!fi REVISIC1:
- Added. "UO2" deleted.
SUPERSEDES: Pace 92 uSNRC APPROVAL REFERE! ICE DATE 4/30/75 REVISICN 0
B2000( & WILWX OIP/M, CaiERCIAL NUCLEAR REL PLAff USNRC LICENSE SNt+-ll68 DOCFEr 70-1201 SECTICri V CONDITIONS 8.0 Technical Scecifications - Radiation Safety 8.4 Personnel Monitorina and Contamination Control 8.4.4 Personnel Contamination 8.4.4.1 a. Continued - Survey hands and decantaminate as necessary to assurc that contagination levels do not exceed 3. O DPM/ 50 cm, or equivalent. - Survey anti-contamination clothing, if con-tamination levels age more than 350 but less *** than 3500 DPM/ 50 cm, or equivalent, don clean protective clothing over the contami-nated garments. Anti-contamina tion clothing with contgmination levels greater than 3500 DPM/ 50 cm, or equivalent, shall be removed. - Remove shoe covers worn in controlled areas, or place clean shoe covers over contaminated shoes. b. Other than as noted above, use of anti-contami-nated clothing for temporary or non-routine activities shall be as stated in Paragraph 8.4.6 of this section. 8.4.5 Surface Contamination Control The following criteria shall be applied to the monitoring and control of surface contamination. The levels shotin belew represent contamination levels which, when exceeded,will cause clean-up activities of the affected area to be initiated.*** Area Actic9 leval Survey Frecueacy 1. Contro tellet y,$oc ynfjeg enfl Weekly
- *1 g m agte.
,[- C* d Total 50,000 CF:V SJ cm rocc(hotarea) 2. Change Roor3 Removable-SM C.a**nM 2 (Intenr.ediate) ' Total 2.000 OPM/ 50 cm
- d '*
Daily
- kten scwder ;r: cessing c:e ations (not includir; roi.tfre 4.211ty cer. trol activities) are tetr.; ::ncocted.
In r.o case wili :r.ere be ccre than one week ::etween survey.s. 3. Clean Remvable: 500OPM/50c-l 200 0?*/100 c-(Re afnder f Total Montgiy
- A*
p' ant and office (hee:: cafete f a area) 494 locke-rec:s sna;; 3,,,3cy) DATE / 0/79 REVISIC(i UO. PAGE 94 0 Deleted; Action Levels Revised. CURRENT REVISICN: 2 SUPERSEEEs: PACE C'- USNRC APPROVAL REFERENCE I/ / I DATE REVISICri }}]g ]gg
BABC00( & WILCOX CDPRf(, C0ffERCIAL NUCLEAR REL PIM USNRC LICENSE SNM-ll68 DOCKET 70-]201 SECT!Oj V CONDITIONS 8.0 Technical Soecifications - Radiation Safety 8.4 Personnel Monitorina and Contamination Control 8.4.5 Surface Contamination Control (continued) When general levels in excess of those noted above occur, prompt decontamination will be undertaken. The levels established for Item "1" above are defined for the general area including floors and other exposed surfaces. The contamination levels in hoods and on equipment that directly contacts SNM cannot~ be effectively *** specified. Control shall be maintained through frequent cleanup and limitation of powder accumulation. Tools and equipment removed from controlled areas for uncontrolled use elsewhere in the plant shall not 2 2 exceed 1000 dpm/50 cm total alpha, or 100 dpm/100 cm removable alpha contamination.' Release of equipment or tools at contamination levels exceeding the above requires the specific approval of Health-Safety. In such cases, Health-Safety may impose limitations on use or require other controls to preclude contamination spread. 8.4.6 Health-Safety may approve the performance of non-routine or temporary operations in the manufacturing area outside the controlled area when the operation may result in contamination provided the following controls are established and documented: - The area involved is identified and access controlled. - Necessary area and personnel contamination control and monitoring programs are implemented including use of anti-contamination clothing if needed and air sampling as required. DATE 10/13/7S REVISICri NO. 2 PAGE 95 CURRE F REVISICN: *" U0,3 Deleted; Centamination Levels / Terminology Adjusted. SUPERSEmS: PACE 95 USNRC APPROVAL FEFERENCE DATE 1 la / 7') REVISICt1 1 8 90
Y'. PAQ0( & WILWX COP #f/, C&TERCIAL NUCLEAR FLEL PLM USNRC LICENSE SNM-ll68 DOCKET 70-]201 SECTIO 4 V CONDITIONS 8.0 Technical Specifications - Radiation Safety 8.4 Personnel Monitoring and Contamination Control S.4.6 Continued - During the course of the cperation decontamination will be initiated when general contamination levels in excess of 1000 DPM/100 cm2 (removable) or 2000 DPM/50 cm2 (total) are noted. Following completion of the operation, decontamination to clean area levels specified in Paragraph 8.4.5 shall be accomplished. 8.5 Intentionally left Blank 8.6 Non-Exemot Sealed Source Control 8.6.1 Use of non-exempt sources for training and instrument calibration shall be limited to, or under the direct control of, qualified Health-Safety personnel. 8.6.2 Sources utilized as a functional component of devices 2318 191 DATE 10/13/73 REVISIC1 NO. 1 PAGE 95a CURRENT REVIS!Cel: ***S.5 Celeted--Redundant; Contamination Levels /Terminolccv Revised. SUPERSEDES: PAC,g 95a USNRC AFPROVAL REFEREi4CE DATE I/24/78 REVISIQi 0
BABC00( & WILCDX C& PAT /, CaiERCIAL NUCEAR FLEL PLM USNRC LICEf4SE SNM-ll68 DOCKET 70-lN_1 SECTICN V C0f1DITIC'!S 8.0 Technical Soecifications - Radiation Safety 8.8 Fire Protection 8.8.2 Continued "!!etl-X" are available in areas where metal fires may occur. Sprinkler systems and other water-type ex-tinguishing systems are not installed in moderation controlled areas. 8.8.3 As part of its emergency program, the CNFP organization shall include a " fire brigade" staffed by qualified personnel, familiar with basic fire fighting techniques, the equipment available for their innediate use on-site, and the nuclear safety and health physics con-siderations that are involved. Fire brigade members shall be retrained at least annually. 8.8.4 Deleted. 8.8.5 Areas under moderation controls shall be prominently posted and approved fire fighting techniques defined. Deviations from the posted techniques can be approved only by one of the following members of the CNFP Emer-gency Team: 2318 192 12718 DATE 10/13/79 REVISICf1 NO. 1 PAGE 101 CURRENT REVISICf1:
- S.S.2 Deleted - Redund m '
SUPERSEEES: PAGE IN ENRC AFFROVAL REFERENCE DATE 4/30/75 REVISICt1 0 _}}