ML19309A733
| ML19309A733 | |
| Person / Time | |
|---|---|
| Site: | 07001201 |
| Issue date: | 01/09/1980 |
| From: | Zeff D BABCOCK & WILCOX CO. |
| To: | |
| References | |
| 15168, NUDOCS 8004010157 | |
| Download: ML19309A733 (35) | |
Text
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f3DR 70 -l2 o i Babcock &Wilcox c -nm ca m ociea,ru e eni P.O. Box 800. Lynchburg Va. 24505 Telephone: (804) 384.5111 January 9, 1980 Mr. L. C. Rouse, Chief Fuel Processing 6 Fabrication Branch Office of Nucicar b!aterial Safety and Safeguards Division of Fuel Cycle 6 blaterial Safety United States Nuclear Regulatory Commission Washington, D. C.
20555
REFERENCES:
(1) SNht-1168, Docket 70-1201
Subject:
Ah!ENDb!ENT APPLICATION NO. 79-3 Gentlemen:
The Babcock 6 Wilcox Company Commercial Nuclear Fuel Plant requests amendment of License Number SNFI-1168 as previously discussed with members of your staff.
Attached to this letter is BGW Check Number 23345 for
$1,400 as required by 10 CFR 170 for a minor safety amendment to our SN51 license.
The areas addressed in the amendment request are as follows:
- Revision of action IcVels for air effluent releases to unrestricted areas.
Revised levels are needed in order to allow for routine localized operations involving pellets outside the controlled area.
A series of detailed controlled experiments were performed to deter-mine appropriate controls for maintaining airborne concentrations in such areas as low as reasonably achievabic.
During the two calendar quarters that the experiments were performed, extensive. surface con-tamination surveys were conducted in the area.
Results of these surveys revealed no new health and safety concerns since existing controls and limits for the uncontrolled area were not excceded.
Based on these tests and experiments, the most effective air-capture technique was selected for implementation, and we are therefore re-questing a corresponding revision of the action IcVel.
- Revision of action levcis for liquid effluent releases.
Routine releases from the CNFP retention tank had been at 1cvols requiring approval and/or further dilution prior to release.
Since aur license currently requires investigation of probabic cause and impicmentation 8004010157 15183 The Babcock & Wilecx Company / EstabHshed 1867
'i.
o Babcocka3Nilcox hbr. L. C. Rouse
.Page 2 January 9, 1980
- Revision of action levels for liquid effluent releases. (cont'd) of corrective action at 6% of MPC and above, an ultrafiltration system was purchased and installed at a cost of approximately $30,000 to remove uranium particulate from certain liquid streams entering the retention tank.
Subsequent to insta11atioc. of the system, we have closely monitored the " permeate" or " cleaned" water tank of the ultrafilter and found contamination levels at 20% t f MPC or below.
While this permeate water is usually diluted by cther streams in the retention tank, at certain times the ultrafilter nay be the only input to the retention tanks.
Thus, we are requesting revision of the CNFP license to speci-fy action at levels above 20% of MPC only.
- Revision of isolation criteria bett.cen fuel assembly storage array and fuel assembly shipping container loiling area.
Sections III and V of our license presently address the fuel assembly processing and storage area as a single array, with other arr ys isolated from it.
Due to limited existing floorspace and the potential need to expand the fuel assembly storage rack array, the fuel assembly shipping container load-ing area will need to be closer than 12 feat from the assembly storage rack.
Since Section 7.10.2 of criticality demonstrat1on of our license states that rotation of one planar array 90 degrees frca the vertical to pre-sent its edge to the face of the opposing array s(gnificantly reduces interaction probabilitit., the loading of assembliis into shipping con-tainers not closer than 38 inches from the nearest a.sembly storage rack is conservative.
Thus, other activities equally ar less reactive than the storage rack may be conducted at a ninumum of 18 inches from the' storage rack.
- Reorganization within the Safety, Licensing, and Safeguards.and Nuclear Materials Control Group.
The Safety, Licensing, and Y feguards Group is now known as IIcalth-Safety and Licensing and Nuclear Vaterials Control is now called Safeguards.
The Safeguards Group still mcintains the NMC function and Licensing and llealth-Safety matters are still controlled under the !!calth-Safety and Licensing Group.
The revisjon index indicates which pages were revised to accommodate the organiza-tional change.
Babcock &Wilcox i
Mr. L. C. Rouse Page 3 January 9, 1980
- Three wording changes for clarity.
Pages 2, 40, and 94 in Section V have minor wording changes to clarify interpretation.
If you or members of your staff have any questions regarding the above iteins, please feel free to call me or J. P. Watters at (804) 384-5111.
Sincerely, BABC0CK 6 WILCOX COMPAW COW!ERCIAL NUCLEAR FUEL PLANT t
7 D. W. Zeff, M l
Health-Safet9an,gerd Licensing
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DWZ:cmm Attachment cc:
J. P. Watters (w/ attcchment)
W. F. Heer (w/o attachment)
D. R. Cox (w/o attachment) l l
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SilM-1168 DOCKET 70-1201 REVISION INDEX AMENDMENT APPLICATION N0.
79-3 PAGE 1 of 2 INDEX DATE:
1-9-80 Section Remove Add Reason
,Page Revision Date Page Revision Date III 5
0 4/30/75 5
1 1/9/80 Change to IIcalth-Safety and Licensing III 114 0
4/30/75 114 1
1/9/80 Change to Health-Safety and Licensing III 123 0
9/19/75 123 1
1/9/80 Change to !!calth-Safety and Licensing III 182 0
4/30/75 182 1
1/9/80 Revised Interaction III 194 - 195 0
4/30/75 194 - 195 1
1/9/80 Revised Interaction IV 13 - 14 0
4/30/75 13 - 14 1
1/9/80 Change to ifcalth-Safety and Licensing IV 38 0
4/30/75 38 1
1/9/80 Change to !!calth-Safety and Licensing IV 40 0
9/19/75 40 1
1/9/80 Change to !!calth-Safety and Licensing
'V 2
1 8/11/78 2
2 1/9/80 Change UO to Uranium 2
O x ido V
11 - 12 0
4/30/75 11 - 12 1
1/9/80 Change to IIcalth-Safety and Licensing V
13 0
9/19/75 13 1
1/9/80 Change to IIcalth-Safety and Licensing V
14 0
12/01/75 14 1
1/9/80 Change to !!calth-Safety and Licensing V
17 0
4/30/75 17 1
1/4/80 Change to llcalth-Safety and Licensing V
24 - 25 0
4/30/75 24 - 25 1
1/9/80 Change to !!calth-Safety and Licensing V
26 1
10/13/78 26 2
1/9/80 Change to !!calth-Safety and Licensing V
27 0
9/19/75 27 1
1/9/80 Change to llcalth-Safety and Licensing V
32 0
4/30/75 32 1
1/9/80 Change to ifcalth-Safety and Licensing V
34 - 35 0
12/01/75 34 - 35 1
1/9/80 Change to IIcalth-Safety l
and L.icensing l
V 36 0
9/19/75 36 1
1/9/80 Change to llcalth-Safety and Licensing V
40 0
12/01/75 40 1
1/9/80 Change UO t Uranium 2
Oxido
SilM-ll68
~
DOCKET 70-1201 REVISIOlt IllDEX
- 4EllDMEllT APPLICATI0ft fl0.
79-3 PAGE 2 of 2 INDEX DATE:
1-9-80 Section Remove Add Reason Page Revision Date Page Revision Date r
V 69 0
12/01/75.
69 1
1/9/80 Revised Interaction
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V 76 0
4/30/75 76 1
1/9/80 Air Action Level Change V
77 0
4/30/75 77 1
1/9/80 h'ater Action Level Change V
84 0'
4/30/75-84 1
1/9/80 Change to Health-Safety and Licensing V
94 4
5/25/79 94 5
1/9/80 Add "*" for clarity e
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BAB000( & WILWX CDP #1Y, CGtERCIAL NUCLEAR REL PLAVf USNRCLICENSESNM-ll68 DOCKET 70-L?01 III fiUCLEAR SAFETY ANALYSIS SECTIOl 7.T' General Requirements 7.1.4 Continued modifications to product specifications or design when re-lating to nuclear safety must be reviewed and approved by the Manager, Health-Safety and Licensing and the Safety Board, prior to initiation.
In those cases where proposed modifications may involve a change in the bases on which nuclear criticality safety was originally assessed, a technical evaluation by the nuclear criticality safety function will be initiated by the Manager, Health-Safety and Licensing.
The nuclear criticality safety function noted above is a separate component within the corporate structure and thus is organizationally independent of the CNFP, with no interest in plant operations, other than the nuclear criticality safety aspects.
By established procedure, analyses performed by the nuclear criticality safety function involving potential changes in the bases upon which plant safety was originally assessed are reviewed for appropriateness of method, validity of assumption, and accuracy of the results prior to application in the plant by a qualified individual not involved in the ori-ginal evaluation.
Qualifications of the evaluator and re-viewers are given in Section V.
DATE 1/9/80 REVISIQ1 NO.
1 PAGE 5
l CURREffr REVISION:
SUPERSEDES: PAGE 5
ENRC APPROVAL REFERENCE DATE 4/30/75 REVISlat 0
BAB000( & WILQX COWMY, CHERCIAL NUCLEAR REL PIM USNRC LICENSE SNM-ll68 IOCKET 70.1201 III NUCLEAR SAFETY ANALYSIS - PELLETIZItiG SECTIQ1 7.4-Pelletizing 7.4.4 Moderation Control 7.4.4.2 Unit Moderation Control Criteria ADMItilSTRATIVE CONTROLS (cont'd) 1.
accuracy, by a second individual at least as well qualified as the one per-fonning the initial evaluation. Further processing of the SilM is prohibited until the above approval is obtained.
The following points are typical of those in the process which require dual signature:
- Initial moisture analysis of as-received powder
- Prior to addition of powder lubricant to blender or transport container
- Prior to addition of re-cycle material to a moderation controlled unit, if the ma-terial has been exposed to an uncontrolled environment.
Adherence to the above procedure will be audited by the Health-Safety Supervisor or the Manager, Health-Safety and Licensing at least' monthly.
2.
The addition of powder lubricants to the blender or transport container has been shown to be safe under credible accident conditions. However, in DATE 1/9/80 REVISION No, 1
PAGE 114 CURRENT REVISION:
SUPERSEDES: PAGE 114 USNRC APPROVAL REFERENCE 0
DATE 4/30/75 REVISION I M
BABC00( & WILWX C&P##, CatERCIAL NUCLFAR REL PIM USNRC LICENSE SNii-ll68 DOCKET 70.1201 SECTIO 4 III NUCLEAR SAFETY ANALYSIS - PELLETIZING 7.4' Pelletizing 7.4.4 l4oderation Control 7.4.4.3 Unit Area f4cderation Control (continued) over moderation controlled units or arrays, the overhead shielding essentially serves as a technique for providing "second order" protection. The shield-ing is not required to protect the array from accidental moderation due to pipe rupture since that contingency has been provided for through baffling and doubid containment.
PERIODIC EVALUATI0tl 0F I40DERATION CONTROL EQUIP 4ENT The routine calibration and standards analysis pre-gram for moderation control instrumentation has been described previously.
While safety within an operational area is the responsibility of the area supervisor, Health-Safety conducts routine system audits to verify that all activities are conducted safely and in accord with license conditions.
Additional nuclear safety and health physics audits are conducted by qualified B&W personnel, from outside the CNFP, at least quarterly.
At least monthly, a detailed inspection will be con-ducted by the f4anager, Health-Safety and Licensing, or his qualified designee.
Inspection scope, l
DATE 1/9/80 REVISIQ1 NO.
1 PAGE 123 l
1 l
CURRENT REVISION' SUPERSEDES: PAGE 123 US!!RC APPROVAL REFERENCE DATE 9/19/75 0
REVISIQ1
BABC00( & WIL(DX C&P##, CORERCIAL NUCLEAR REL PlM USNRC LICENSE SNft-ll68 DOCKET 70-1201 SECTION III NUCLEAR SAFETY ANALYSIS 7.'9 Fuel Assembly Processing 7.9.1 The nuclear safety of an individual assembly is discussed in Part 7.10.
The assembly operation in which fuel rods are loaded into the final configuration is safe under moderation control conditions, since the U-235 enrichment is less than 5% by re-ference to TID 7016 and TID 7028 Figure 17'.
The assembly area is a completely closed room of concrete block construction with a metal deck roof topped by poured concrete. There is no sprinkler system in the assembly room.
The restriction of con-tainers to one gallon size effectively prevents the contents from meaningful moderation in the event of spill.
Once the piping is protected from accidental rupture, the joints become the areas most susceptible to failure.
Baffling of the joints to control the direction of water flow provides assurance that the assemblies in process cannot become moderated.
Some amounts of polyethylene type paterial may be in the area for dust pro-tection of equipment, or as pre-assembly protective wrappers on components.
Precluding the polyethylene from interspersion among fuel rods other than when in the 4.0" slab, completes the controls against moderation and obviates the need for further consideration of interaction between assemblies.
7.9.2 For the purpose of safety of the array, fuel assemblies process-ing, storage, and loading into shipping containers are considered as one array which is safe as described in Part 7.10.2.
I DATE 1/9/80 REVISIOJ No, 1
PAGE 182 CURRENT REVISION:
SUPERSEDES: PAGE 182 LENRC APPROVAL REFERENCE DATE 4/30/75 REVISIG1-0 1
Bl0000( 8 WILODX WWW, C&ERCIAL NUCLEAR REL PUNT USNRC LICENSE SNM-ll68 DOCKET 70-1201 III fiUCLEAR SAFETY ANALYSIS SECTIO 4 7.'10 Fuel Assembly Storage 7.10.1 Continued In teraction:
The 21 x 38 inch c/c array is the more reactive.
Since it has been assumed infinite in the x-y plane, its calculated k-effective is higher than that of the combination of the two arrays. The separa-tion between the two arrays is greater than 38 inches so that no third type of array is created.
Therefore, values obtainef for the infinite 38 x 21 inch c/c array may be conservatively used for the actual combination.
For the maximum credible degree of interspersed moderation, k-effective is less than 0.4.
Similarly, all other arrays not isolated from the assembly storage array are also free draining and cannot accumulate low density moderator.
Since they are also low enriched uranium
( < 4 weight %) the same principle applies--the actual combination of arrays is less reactive than the infinite array of 38 x 21 inch c/c assemblies.
7.10.2 Section III, Part 7.10.1 discusses the nuc1 car safety of an infinite array of parallel fuel assemblies stored in 21 inch by 38 inch or 36 inch center-to-center spacing in a vertical con-figuration.
Part 7.9 validates the nuclear safety of fuel as-sembly processing.
fluclear interaction between fuel assembly processing and assembly storage, and also between assembly storage *"*
and assembly shipping container loading is limited to less than the interaction within the storage array alone due to geometric relationship.
Since a planar array of assemblies on 21 inch centers is subcritical for credible interspersed moderation conditions (Part 7.10.1), the planar array in the assembly DATE 1/9/80 REVIS!Q1f40.
1 PAGE 194 CURRENf REVISIQ4:
SUPERSEDES: PAGE 194 LGt!RC APPROVAL REFEREf1CE DATE 4/30/75 REVISIQ1 0
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e Bl&DO( & WILGM C&P/M, CMERCI/L NUCLEAR REL PIM USNRC LICENSE SNM-ll68 DOCKER 70.1201 SECTIOJ III NUCLEAR SAFETY ANALYSIS
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I 7.10 Fuel Assembly Storage 7.10.2 Continued processing area, where four foot separation is maintained, is likewise subcritical.
Two or more planar. arrays have been shown to be subcritical in combination when center-to-center distance is not less than 38 inches, and with major faces opposed, the situation resulting from consideration of an infinite storage array.
Rotation of one planar array 90 degrees from the vertical to present its edge to the face of the opposing array, while main-taining the specified separative distances, significantly reduces interaction probabilities by presenting less " effective" surface area for interaction thus,.in effect, extending the effective separation distance. This demonstration accurately reflects conditions at the junction of the fuel assembly pro-cessing and storage area, and also at the junction of the assem-bly storage and shipping container loading area.
Under normal' operating conditions the activities will be con-l ducted in the absence of any significant moderation; under i
1
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l such conditions, fuel of 4 wt,% enrichment and less cannot be made critical.
DATE 1/9/80 REVISION No.
I PAGE 195 CURRENf RFVISION:
SUPERSEDES: EAbE 195 USNRC APPROVAL REFERENCE DATE 4/30/75 REVISIQ1 0
i BAB000( & WILQX C&PhiY, C&ERCIAL NUCLENT FLEL PUtlT WNRC LICENSE SNitll68 DOCKER 70-3201 SECTIGi IV llEALTH PHYSICS 8.2' Cont'd limited to the following (10 CFR 20.101)
- Whole body, trunk and head, active blood forming organs, lens of eyes, and gonads 1.25 Rem / cal. qtr.
- Hands and forearms, feet and ankles 18.75 Rem / cal.~qtr.
- Skin of whole body 7.50 Rem / cal. qtr.
However, AEC Form 4 documentation is provided for each permanent employee and in specific cases the Manager, Health-Safety and Licensing and the Plant Manager may authorize exposures greater than the above provided 10 CFR 20.101 (b) and 20.102 requirements are sa'tisfied.
In these cases, the "maximu.1 permissible exposure" is taken to be:
- 3 rem per calendar quarter to the whole body,
- 5 rem per calendar year to the whole body,
- And, accumulated dose is < 5(N-18) rem where N is the individual's age at last birthday.
8.3 Exposure of employees to airborne contaminants is limited by the concentrations set forth in 10 CFR 20, Appendix B, Table 1 averaged over 40 working hours in any 7 consecutive days'.
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l DATE 1/9/80 pgyIsIat go, 1
PAGE 13 CURRENT REVISION:
SUPERSEDES: PAGE 13 USNRC APPROVAL REFERENCE MTE 4/30/75 0
REVISIQ1
BABCDCK & WILODX C&P/M, CMERCIAL NUCLEAR REL PUNr USNRC LICENSE SNM-ll68 DOCKET 70-1201 IV HEALTH PHYSICS ggg 8.3' Cont'd Permissible exposure levels are adjusted based on the actual time the individual is present in the area. Additionally, respirators may be used and exposure results adjusted according to the protection factors defined in Paragraph,12 of this section, and in Section V.
In actual practice exposure to airborne activity is maintained as low as practicable by engineered or other controls and a design / operational criteria of < 25% of Appendix B concentration levels is applied.
8.4 Minors (persons under 18 years of age) are limited to doses not greater than 10% of those specified in 8.2.
Exposure of minors to airborne contaminants will not exceed the levels contained in Appendix B, Table II of 10 CFR 20. Minors are not employed at CNFP as a matter of policy.
8.5 Exposure of visitors in all cases is limited to 10% of the levels specified for occupational exposure in 10 CFR 20, unless documen-tation of the radiation exposure history is available.
8.6 Eating and smoking are prohibited in controlled areas.
Smoking may be permitted in change rooms if assurance of adequate radiological hygiene practices are maintained. At the discre-tion of the Manager, Health-Safety and Licensing drinking fountains may be located within cont.'lled areas where non-transportable materials are being processed.
Drinking facilities are provided in change areas.
DATE 1/9/80 REVISION No.
I PAGE I4 CURREfE REVISION:
SUPERSEDES: PAGE 14 USNRC APPROVAL REFERENCE DATE 4/30/75 pgygsta; O
i.
BABC00( & WILGX C&PM/, C&ERCIAL NUCLEAR REL PIM USNRC LICENSE SNM-ll68 DOCKET 704201 SECTIO 4 IV llEALTil PHYSICS 16.' Cont'd 16.2 In addition to the above, logs are established and maintained to document all routine monitoring activities including:
- Air sampling
- Radioactive Material shipments
- Bioassay and Environmental samples
- Source control and leak tests
- Smear and penetrating radiation surveys
- Equipment history, use and calibration
- Pre-operational and operational evaluations
- First aid (per OSHA requirements)
- Contaminated air handling system inspections
- Health physics, Nuclear safety, and industrial safety inspections and audits.
16.3 In-plant records, and reports submitted to regulatory bodies, are reviewed periodically by the Manager, Health-Safety and Licensing ***
and in the course of routine audits prodived by other B&W components.
Record retention requirements are in accord with 10 CFR 20.
16.4 Where applicable, all records are maintained using terminology equivalent to 10 CFR 20.
Where because of type of sample or the analytical technique employed, other units are more descriptive or useful, Health-Safety may maintain records in the units most appropriate.
Typically, unit terminology will be as noted below:
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38 DATE 1/9/80
'REVISIO1 NO.
1 PAGE CURRENT REVISION:
SUPERSEDES: PAGE 38 USNRC APPROVAL REFERENCE 4/30/75 0
DATE REWSIm
BABC00( & WIUDX C@PfM, C0FERCIfi NUCLEAR REL PIM USNRC LICEf!SE SNitll68 DOCKET 70-1201 IV HEALTH PHYSICS SECTIOl 17.' Continued posting plant entrances with signs incorporating the radiation symbol and the following warning:
CAUTION RADI0 ACTIVE l%TERIALS ANY AREA OR CONTAINER WITHIN THIS PLANT IMY CONTAIN RADI0 ACTIVE IMTERIAL.
This sytem has been applied effectively at the CNFP for the past four years.
17.1 Personnel access to the plant, and supporting buildings is con-trolled through a guard station which is constantly manned. The plant area is surrounded by a chainlink fence, and all access points, other than the main gate, are locked when not in use.
Key control is the responsibility of the plant engineer.
For purposes of plant and personnel protection and access control, the CNFP " restricted area" is considered to consist of all areas within the perimeter fence.
In specific instances, the Manager, Health-Safety and Licensing may modify the restricted area as defined above in order to accommodate unusual or temporary conditions.
17.2
" Controlled areas" within the CNFP are established as neces-sary to assist in the protection'of personnel or in the control of contamination spread.
Criteria specified in 10 CFR 20.202 and 20.203 are applied to the determination for the necessity of establishing controlled areas.
L DATE 1/9/80 REVISIOJ NO, 1
PAGE 40 CURRENT REVISIQl:
l SUPERSEDES: PAGE 40 LENRC APPROVAL REFERENCE DATE 9/19/75 pgyggygg 0
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BABUXX a WIUDX C0FP/W, C0FERCIAL NUClfAR REL PU#f USNRC LICENSE SfM-ll68 DOCKET 70-3201 SECTIO 4 V CONDITIONS 2.0 Authorized Activities are as follows:
2.1 Fabrication of nuclear power fuel assemblies using Uranium 0xide powder ***
or pellets as starting material, through pellet production, fuel assembly fabrication, packaging and delivery of fuel assemblies to a carrier fo: transport.
2.2 Disposal of various solid, liquid, and airborne wastes resulting from the authorized activities, excluding on-site burial.
2.3 Laboratory operations such as but not limited to chemical analysis, metallographic analysis and testing.
2.4 Storage of nuclear materials in various forms and facilities appropriate to safety.
2.5 Activities of a process and product development nature for those technologies authorized by SUM 1168.
2.6 Maintenance of the facilities and equipment under adequate control to assure safety.
2.7 Possession of authorized by-product, source, and special nuclear materials in packages approved pursuant to 10 CFR 71 for the purpose of delivery to a carrier for transport, and in private carriage between NRC licensed facilities within the United States.
l l'1 ATE 1/9/80 REVISION NO.
2 PAGE 2
CURRENT REVISIG4:
SUPERSEDES: PAGE 2
USNRC APPROVAL REFERENCE DATE 8/11/78 I
REVISIQ1
i' B/BC00( & WIUDX COW ##, CORERCIAL NUClFR FLEL PLMf
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USNRC LICENSE SNM-ll68 DOCKET 70-1201 SECTICN V CONDITIONS 6.0' General Specifications - Implementation of the technical specifications will be accomplished through application of the following general specifications.
6.1 Control program Specifications and design criteria for purchased or locally fabrication equipment where nuclear and radiological safety considerations are involved are approved by a knowledgeable representative of Health-Safety and Licensing.
Before being released for production operation, new equipment is tested to assure that safety specifications are satisfied.
Safe geometry equipment will be measured by a knowledgeable person to ascertain that it is of proper dimensions before it is put into service.
Where operational safety is based wholly or in part on the use of electrical or mechanical interlocks, the proper functioning of interlocks shall be verified upon installa-tion and on an annual basis thereafter.
Sintering furnace safety interlocks will be tested prior to each startup.
Routine plant inspections place added emphasis on new operations. No equipment is used after being removed from service until an equipment checkout for continued effectiveness of safety related parameters is performed.
I DATE 1/9/80 REVISION NO.
1 PAGE 11 CURRENT REVISION:
SUPERSEDES: PAGE 11 USNRC APPROVAL REFERENCE DATE 4/$/75 PEVISION 0
BABC00( a WILWX C0FPAW, COWERCIAL NUCLEAR REL PUM USNRC LICENSE St#i-ll68 DOCKET 70-3201 SECTIai V CONDITIONS 6.0' General Specifications 6.1 Radiation instruments are calibrated at the frequency indicated in Section V, Part 8.3.4.
Calibration is performed by Health-Safety personnel.
Ventilation, containment, and air cleaning equiement is routinely inspected by Health-Safety personnel to assure continued effectiveness and compliance with license specifications.
These inspections are recorded and a periodic review is per-formed by the Health-Safety Supervisor, and Manager, Health-Safety and Licensing.
l The Health-Safety Supervisor or qualified designee is responsible for control of the air sampling and contamination survey pro-gram, and evaluates sample results on a daily basis.
Sample l
and survey records are periodically reviewed by the Manager, l
Health-Safety and Licensing.
Instances of inadequate control of contamination or air i
effluent levels, and of equipment operation not in accordance i
with established nuclear and radiation safety and license l
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DATE 1/9/80 REVISIQ1 NO.
1 PAGE 12 CURRENT REVISION:
SUPERSEDES: PAGE-12 USNRC APPROVAL REFERENCE 0
DATE 4/30/75 REVISIG4
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BABC00( a WILQX COPP/M, CORERCIAL NUCLEAR REL PMK USNRC LI&NSE SNti-ll68 DOCKER 70-1201 V CONDITIONS sectral 6.0' General Specifications 6.1 Control Program (continued) parameters are reviewed by either the Health-Safety Supervisor or the Manager, Health-Safety and Licensing who may order the operation suspended if timely compliance cannot be assured.
Disciplinary action can be recommended if warranted.
SAFETY REVIEW BOARD A Safety Review Board shall be established to consider the effect of new or revised facilities, to analyze equipment and processes' involving radioactive materials and to evaluate the continuing effectiveness of established controls anr' safeguards, including maintenance of "as low as practicable" criteria.
In the case of minor procedural changes and where existing safety practice remains unchanged, the Manager, Health-Safety and Licensing may determine that Boar'd review is not necessary.
The Board shall be chaired by the Manager, Health-Safety and Licensing, and the permanent membership shall consist of representatives of CNFP operational and technical management. - Technical representatives of consulting organi-zation shall be included as necessary (LRC, Nuclear Criticality Safety Group). The Manager, Health-Safety and Licensing reviews all requests for changes in process and equipment I
which involve radioactive material and determines if Board DATE 1/9/80 REVISION NO.
1 PAGE 13 CURRENT REVISION: ***
SUPERSEDES: PAGE 13.
USNRC APPROVAL REFERENCE DATE 9/19/75 REVISION 0
BECDO( & WILCOX COPPM, CMERCIAL NUCLEAR REL PLM USNRC LICENSE SNtt-ll68 DOCKER 70.1201
,SECTION V CONDITIONS 6.0' General Specifications 6.1 SAFETYREVIEWBOARD(continued) review is necessary. The Board will not approve any new or revised facility, equipment, or process until it has received written determination (from the 14anager, Health-Safety and Licensing) for the necessity of a nuclear safety evaluation and/or license amendment, and has reviewed where required, a properly documented nuclear safety evalua-tion or approved license amendment.
Board meetings may be convened at the discretion of the Manager, Health-Safety and Licensing, but shall be held at least quarterly.
Records of Safety Review Board proceedings, including support-ing calculations and approvals, will be retained for at least six months after the completion or termination of the subject activity.
RECORDS Operating logs and records of reviews, tests, inspections, personnel qualifications, and audits shall be maintained on file for review by B&W management and regulatory agencies.
Duration of record retention is based on existing federal regulations, license specifications, customer requirements and prudent management practices.
6.2 EMERGENCY PLAN An emergency plan and implementing procedures for coping with radiation emergencies shall be maintained.
Emergency pro-cedure content shall include:
a.
An or2anization for coping with radiation emergencies, including definition and assignment of authorities, responsibilities, duties, qualification, and training requirements.
Notification steps for "in-house" personnel and federal, state, and local authorities shall be included.
b.
A list of employees and other individuals, in addition to those assigned to the emergency organization, who have special j
qualifications for coping with emergency conditions.
l c.
The actions planned to protect the health and safety of l
individuals and to prevent damage to property both within j
ar.d outside the site boundary in the event of credible l
emergencies, including the means for determining the mag-nitude of the release of radioactive materials, guidelines for evaluating the need for notification of local, state, and Federal agencies, and the type and extent of protec-tive action to be taken within and outside the site boundary.
DATE 1/9/80 REVISIW NO.
I PAGE I4 CURRENT REVISION:
SUPERSEDES: PAGE I4 LENRC APPROVAL REFEFGNCE DATE 12/1/75 O
REVISION
BABC00( & WIL(DX @P/M, CMERCIAL NUCLEAR REL PUM USNRC LICENSE SNM-ll68 DOCVEr 70.1201 SECTIa1 V CONDITIONS
~
6.0 General Specifications
+
6.3 Personnel Training (continued) is maintained by Health-Safety.
The training of personnel assigned to the Health-Safety func-tion is the responsibility of the fianager, Health-Safety and Licensing.
The extent and depth of the training is based on the specific job assignment involved.
Heal th-Safety monitoring personnel shall reveive a combination of formal and "on-the-job" training such that they can successfully demonstrate their proficiency in basic nuclear and radiation j
\\
physics monitoring and control techniques and regul? tory re-quirements.
The degree of proficiency will be determined on an individual basis by practical and written examinations, the results of which will be maintained on file by Health-Safety.
Member of the CNFP emergency monitoring team shall be trained by Health-Safety in the proper technique of accident control.
Training program content shall be as outlined in 6.3.1 through 6.3.3.
DATE 1/9/80 REVISIQ4 NO.
I PAGE 17 CURRENT REVISION:
SUPERSEDES: PAGE 17 USNRC APPROVAL REFERENCE 0
DATE 4/30/75 REVISIG1
BABC00( a WILCDX C@P/M, COTERCIAL NUClfAR REL PUM WNRC LICENSE SNM-ll68 DOCKET 70.1201 SECTIQ1 V C0f1DITI0t1S 6.0' General Specifications 6.5 EnrichmentControl(continued) re-establishment of the identity.
6.6 Procedure Control Written procedures for the conduct of specific operations in-cluding maintenance and development work within the plant are prepared by the functional component responsible for that activity and are reviewed and approved by appropriate respresentatives of plant management!.
Where the activity under consideration involves SilM or radioactive materials, specific written approval by the Manager, Health-Safety and Licensing or his designee is required prior to implementa-tion of the procedure. Adherence to procedure is required of all personnel.
Procedures for operations where nuclear and radiological safety are involved shall include specific
- aference to applicable safety requirements.
Procedure and format shall be such that operations are clearly detailed and specific directions are provided for operation under both normal and abnormal conditions.
Deviation from written pro-cedures for the handling of Stim shall be approved by the Manager, Health-Safety and Licensing, or his qualified l
alternate.
Procedural control of activities at the CilFP are l
categorized as follows:
DATE 1/9/80 REVISION NO.
1 PAGE 24
~***
CURRBIT REVISION:
SUPERSEDES: PAGE 24 WNRC APPROVAL REFEREf1CE 0'
DATE 4/30/75 REVISIQ1
.1
B/BC00( 8 WILCDX CDP /W, C0ffERCIAL tRICLEAR REL PUM LGNRC LICEtlSE SNft-ll68 DOCVEr 70-1201 V CONDITI0tlS SECTICt1 6.0' General Specifications 6.6 Procedure Control (continued)
- Procedures developed by ifealth-Safety specifying the method by which safety related functions are to be accom-plished.
Such procedures may be for internal Health-Safety use or may be intended for general distribution to affected individuals within other components.
Health-Safety procc-dures shall be approved in writing by the Plant Manager, the Manager, Health-Safety and Licensing, and other members of plant management / supervision if it is determined by Health-Safety that their area of responsibility is affected by the procedure.
- Safeguards procedures providing the tech-niques for the accountability and measurement of SNM.
Such procedures will be approved in writing by the Manager, Health ***
Safety and Licensing and by affected plant management.
- Manufacturing, Quality Control, and selected Maintenance /
Engineering procedures, where nuclear or radiological safety, license conditions, or regulatory requirements are involved require prior approval by the Manager, Health-Safety and l
Licensing, as well as approval by affected members of plant management.
l I
DATE 1/9/80 REVISICJtl0.
PAGE 25 CURREtiT REVISION:
SUPERSEDES: PAGE 25 UStlRC APPROVAL REFEREllCE DATE 4/30/75 REVISIg; 0
BMC00( a WIL(DX C&P/M, CGERCIAL NUClfAR REL PLM USNRC LICENSE SNM-3168 DOCKER 70-1201 SECTlal 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, Health-Safety and Licensing.
Health-Safety supervision shall conduct, at least weekly, a formal 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.
DATE 1/9/80 REVISION N0, 2 PAGE 26 CURRENT REVISION:
SUPERSEDES: PAGE 26 USNRC APPROVAL REFERENCE DATE 10/13/78 REVISION I
IXOWUN Cu HILUJA W f flil / WTLILilt 61ULLLIU\\ i ULL ruiti
~
USNRC LI NSE Snit-ll68 DOCKER 70-3201 i
SECTIm V C0tlDITIONS
- 6. 'O General Specifications 6.7 Audits and Quality Assurance (continued)
Health-Safety personnel shall, as part of their routine duties, conduct informal daily audits of plant activities.
Independent auditors shall conduct inspections as follows:
- Nuclear safety
- quarterly
- Health physics
- quarterly The audit program shall include physical inspections and records review for the industrial, nuclear, and radiological safety elements of plant activities including:
- Effectiveness of procedural controls impacting on operational safety parameters
- Review of equipment design, layout, and operation for
~
continued compliance with safety design criteria
- Audit of operating records, where such records provide a means of verifying procedural compliance with safety specifications
- Periodic review and evaluation of contamination survey data Independent auditors' reports shall be submitted to the Plant Manager. The Manager, Health-Safety and Licensing shall review audit findings for the satisfactory completion of any necessary remedial actions.
Audit findings are reviewed by the Safety Review Board.
i 4
DATE 1/9/80 REVISION NO.
I.
PAGE 27 CLERENT REVISIGN:
SU?ERSEDES:
PAGE 27 LGNRC APPROVAL REFERENCE l
DATE 9/19/75 REVISION 0
uruwuv o.- runwiccu rnirTurrcnunc uuuumvru_x ruiu USNRC LlCEf1SE SNtt-ll68 DOCKET 70.1201 ITIONS SECTIQ1 7.'0 fluclear Safety - Technical Specifications 7.1
_ General Requirements (continued) 7.1.2.1 The flanager, Health-Safety and Licensing, with the concurrence of the LRC fluclear Cri-ticality Safety Group, will determine which arrays are more reactive than those used as a basis for a, b, and c or are best represented as slabs for nuclear interaction purposes.
In these cases, the interaction acceptance cri-teria will be evaluated on an individual basis by the LRC fluclear Criticality Safety Group or it will be the equivalent of eight inches (20.3 cm) of water er as defined in 7.1.2(d)or(e).
7.1.3 Isotopic enrichment of SllM will be verified by shipper's certifications with added assurance by Quality Control statistical overchecks.
l I
DATE 1/9/80 REVISIG1 NO.
I PAGE 32 CURRENT REVISION:
SUPERSEDES:
PAGE 32 USNRC APPROVAL REFERENCE DATE 4/30/75 REVISION 0
DIUWUs 6 WILWA WTh(I, W i Litill i1ULLLM ilLL l'Util USf!RC LICENSE SffM-ll68 DOCKET 70.1201 SECTICtf V CONDITIONS
=
7.d fluclear Safety - Technical Specifications 7.1 General Requirements (continued) 7.1.4.2 proposed modifications shall be demonstrated to satisfy the following criteria to approval:
a.
The license condition specified in this part are not modified.
b.
Huclear safety calculations are performed in a manner equivalent to those originally utilized to demonstrate safety for the area and type of operation under consideration.
Calculations shall consider both individual unit and interaction safety and shall apply the basic assumptions and input parameters used in the original calculations.
c.
Calculated reactivity levels do not increase based on the assumptions and input parameters described below.
7.1.4.3 The review and approval procedure shall be as follows:
a.
A written request shall be submitted to the Manager, Health-Safety and Licensing describing the proposed change.
b.
The proposal will be reviewed by the Manager, Health-Safety and Licensing for content, completeness, and compliance with license DATE 1/9/80 REVISION i0, I
PAGE 34 CURREffT REVISION:
SUPERSEDES:
PAGE 34 LGNRC APPROVAL REFERENCE DATE 12/1/75 REVISION 0
ITTtlLild_ ((ULLUUt t ltL l'Unil
~-
U3NRC LICEt1SE SNM-ll68 DOCI'ET 70-]201 SECTIGJ V C0tlDITI0tlS 7.D Nuclear Safety - Technical Specifications 7.1 General Requirements (continued) 7.1.4.3 Continued b.
specifications and, where nuclear safety evalua-tion is required, forwarded to the fluclear Criticality Safety Function (liCSF).
TheNCSFperformsnecessiirycalcu;ationsinaccord c.
with 7.1.4.2 b and forwards the results in writing to the Manager, llealth-Safety and Licensing.
Evaluations performed by the NCSF are reviewed as specified in Part 5.5 of this section.
d.
The results of the nuclear safety calculations and the recommendations of the NCSF and fianager, 11ealth-Safety and Licensing are reviewed by the Plant Safety Board and the permissibility of the change determined and documented.
c.
Appropriate personnel are notified in writing of the result of the review by the llanager, llealth-Safety and Licensing.
f.
The proposed change will not be implemented until a pre-operational audit (Ref.Section V, Part 6) has been satisfactorily completed.
g.
Correspondence calculations and other material i
will be maintained on file by llealth safety 35 DATE 1/9/80 REVISION NO.
1 PAGE CURREfff REVISION:
SUPERSEDES:
PAGE 35 USNRC APPROVAL REFERENCE I
I 0
_ XdL
_ ____ __ _ REVISION
BABWOC & WILQX C& PAW, C&iERCIAL NUClfAR REL PUffi USNRC LICEllSE Stitt-ll68 DOCKET 70-1201 V CONDITI0flS SECTICf1 7.0 ?!uclear Safety - Technical Specifications 7.1 General Requirements (continued) 7.1. 4. 3 Continued g.
for at least 6 months following termination of the operation.
7.1.4.4 iluclear safety evaluation is not required where the anticipated change does not affect existing nuclear safety controls or bases.
The Manager, Health-Safety and Licensing or his qualified designee determines when a change in product, process, or equipment requires a nuclear criticality evaluation or license amendment.
w l
DATE 1/9/80 REVISION NO.
1 PAGE 36 CURRENT REVISION:
SUPERSEDES:
PAGE 36 USl1RC APPROVAL REFERENCE es a w O fl O 17f f%F"11F f* T #M t a
- v.,,, - m,,,,
m
,_>mi,m,1-m mm, uu,
USNRC LICEf1SE St#t-ll68 DOCKET 70.1201 SECTIQ1 V C0f1DITI0tlS 7.0 Nuclear Safety - Technical Specifications 7.2 Materials Specifications - The following specifications for materials are based on the various criticality control evaluations discussed in Section III and may be considered typical of material processed.
7.2.1 Fuel Pellets Composition:
Uranium Oxide Size:
.600[l diameter
.300 235 Enrichment:
to 4.05%
0 NOTE: An increase from 4.00 to 4.05 w/o does not signi-ficantly affect nuclear safety calculations based on 4 w/o material.
I DATE 1/9/80 REVISION l'O.
PAGE 40 CURRENT REVISION:
SUPERSEDES:
PAGE 40 USNRC APPROVAL REFERENCE DATE 12/1/75 REVISION 0
BABC00( & WILCD$ CUP #N, c0VERCIAL NUCLEAR FLEL PUNT USNRC LI&NSE SNM-ll68 DOCKET 70-1201 f
SECTION V C0tlDITI0flS 7'O Nuclear Safety - Technical Specifications 7.9 Fuel Assembly Processing (cont'd) 7.9.1 Continued accumulation of fuel rods once the fuel rods have been
' removed from the geometric safe 4.0" slab.
i A minimum separation of 4 feet will be maintained be-tween bundle assembly stations, and the equipment will i
be such that liquids cannot be retained in and around the in-process operation.
7.9.2 Assembly processing operations once the fuel rods are restrained in the fuel assembly configuration may be performed without moderation control. The assemblies will be separated by'at least 36" center-to-center.
7.9.3 For the purposes of nuclear interaction control, the funl assembly processing, fuel assembly storage, and fuel assembly shipping container loading areas are con-sidered as a single array with adjacent arrays being no more reactive.
- n*
7.10 Fuel Assembly Storige and Packaging 7.10.1 Fuel assemb' y dust wrappers, if used, will be arranged to permit dr ainage of water from within.
Moderation,
(
such as pol.' ethylene, etc., will not be permitted within the assemblies.
DATE 1/9/80 REVISION NO.
I PAGE 69 CURRENT REVISION:
SUPERSEDES: PAGE 69 USNRC APPROVAL REFERENCE DATE 12/1/75 0
REVISION
vu_m i u_i_ i u ni V on ',
USNRC LICEllSE SNf4-ll68 DOCKET 70-El V CONDITI0ftS SECTIQ1 j
8.0 Technical Specific:
is - Radiation Safety 8.1 Effluents to Unrestricted Areas 8.1.1 Continued c.
The following program shall ba instituted to assure that airborne releases to uncontrolled areas are main-tained as low as practa:ble:
Quarterly Avera50 - % of 10 CFR 20 Appendix B, Table 2 LIMIT ACTION 1 0% of "MPC"
!!one required.
1 11 - 20%
Conduct investigation of system and !
correct if possib1e.*
l 21 - 75%
Visually inspect system. As soon as practicable, conduct efficiency test. Determine and correct cause.**
> 75%
Conduct immediate investigation to determine and correct problem in ventilation system. Maintain addi-tional Health-Safety surveillance of system to monitor contamination levels until problem is corrected.
, end weld operation - flone required end weld operation - l'one required at 1 50% !!PC
- wa
-j Health-Safety shall maintain records of investiga-tions and actions resulting therefrom, d.
Pre-filters serving equipment in the pelletizing area wil,1 be instrumented an:1 monitored as speci-fied in 8.1.la above.
8.1.2 Potentially contaminated liquid effluents will be controlled through a retention tank system with dilution and mixing capabilities and shall be evaluated DATE 1/9/80 REVISIQ1 NO.
1 PAGE 76 CURRENT REVISIGl:
i SUPERSEDES: PAGE 76 USNRC APPROVAL REFERENCE DATE 4/30/75 REVISIQ1 0
/
~~ ~
BABOXX a WILODX COWM/, CODERCIAL NUCLEAR REL PUiff f
USNRC LICENSE SNti-ll68 DOCKER 70-1201 ONDITIONS SECTIOl 1
8.'0 Technical Specifications - Radiation Safety 8.1 Effluents to Unrestricted Areas 8.1.2 Continued for compliance with 10 CFR 20, Appendix B limits prior to release to unrestricted areas.
The following actions, at the contamination levels indicated, will be undertaken in order to maintain contaminated liquid effluent re-leases as low as practicable:
% 10 CFR 20 Apo. B. "itPC" Action
< 20%
No action required.
- u*
21 - 75%
Individual releases authorized
,by Plant Manager or his alter-nate.
> 75%
Discharge prohibited. Waste routed for further treatment or disposal by burial.
Health-Safety shall maintain records of investigations and actions resulting therefrom.
a.
Retention tanks will be inspected monthly for sludge accumulation.
l 1/9/80 DATE REVISIQ1 No.
1 PAGE 77 CURREllT REVISION:
SUPERSEDES: PAGE 77 USNRC APPROVAL REFERENCE DATE 4/30/75 REVISIQ4 0
4
.,_., -.. -, ~ ~.. -.. _
m.,,
USNRC LICENSE Sfiti-ll68 DOCKET 70.1201 V C0flDITI0tlS SECflai 8.0 Technical Specifications - Radiation Safety 8.2 Ventilation, Containment, and Airborne Activity Control (cont'd) 8.2.3 Air Sampling (continued)
The following action points shall apply to the routine air sampling program:
Exposure - % MPC (Time Weighed Average)
Action
< 25%
No action required.
> 25%, < 50%
Report to Manager, Health-Safety and Licensing or designee.
Evaluate operation and contain-ment.
Increase air sample frequency if indicated.
> 50%, < 75%
Report to Manager, Health-Safety and Licensing. ***
Take steps necessary to lower l
airborne activity to acceptable range.
~~
> 75%
Notify Plant Manager.
Terminate operation if > 100% MPC.
!!adify or Install air capture devices.
Increase air sample frequency.
Health-Safety shall maintain records of investigations and corrective actions taken.
The above action levels do not apply to non-routine
~
operations conducted with the cognizance of l
l DATE 1/9/80 REVISION NO.
I PAGE 84 CURRENT REVISION:
SUPERSEDES:
PAos 84 USNRC APPROVAL REFERENCE DATE 4/30/75 REVISION 0
BABWLK & WILWh LU nm, wI thuuu_ uutum ru_t um n..t USNRC LICENSE SNti-ll68 DOCIET 70-1201 SECTIG1 V CONDITIONS l
8.
Technical Specifications - Radiation Safety 8.4 Personnel Monitoring and Contamination Control 8.4.4 Personnel Contamination 8.4.4.1
- a. - Survey anti-contamination clothing, if con-tamination levels are more than 350 but less than 3500 DPM/50 cm2, or equivalent, don clean protective clothing over the contami-nated garments.
Anti-contamination clothing with contamination levels greater than 3500 d
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 temocrary 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 monitoring and control of surface contamination.
The levels shown below represent contamination levels which, when exceeded, will cause clean-up activities of the affected area to be initiated.
Area
[ctionlevel Survey Frecuency
- 1. Controlled tellet ren3 able: 5,000 DM*/IC3 ("
fabricatloa/ rod tota) 50,cuu CriV 53 cn Icadingtt.ng-)
reco(hotarea i
- 2. Change hcra Romvable:
2 533 CM1/100 c Daily *
(Intercediate)
- lotal 2.033 DM;/ 50 car
- Uhen poder processing c;erations (r.ot includir; ratir.e quality cor.*rol activities) are being cond;cted. In r.o case hill there be scre that sne eek bet.een surveys.
2 l
3.
Clean Ker.ovable:
200 t?.M/100 c 2 F.cnthly (Fc.aleder of Total 503 D?M/ 50 cm l
plant and effice area)
- The lunchroom and locker rooms shall be surveyed daily when ***
in use and removable contamination kept as low as reasonably achievable.
DATE 1/9/80 REVISION NO.
5 PAGE 94 CURRENT REVISION:
15168 1 SUPERSEDES:
PAGE,94 USNRC APPROVAL REFERENCE -
~
I L
DATE.
5/25/79 REvtsION 4
]