ML20155H282

Proposed Rule 10CFR71, Transportation Regulations: Compatibility W/Iaea. Rule Would Amend Regulations Re Safe Transportation of Radioactive Matl

ML20155H282
Person / Time
Issue date:	05/23/1988
From:	Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:
References
FRN-53FR21550, RULE-PR-71 PR-880523, NUDOCS 8810190384
Download: ML20155H282 (127)

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- NUCLEAR REGULATORY COMISSION On7nal sent to the C; M4728 lggg }9 Ott,ce of the Federal Register Ag 10 CFR PART 71 for publicaton M ransportation Regulations: Compatibility with s ,y the International Atomic Energy Agency (IAEA)

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AGENCY: Nuclear Regulatory Comission. 000KET fM.1BER PROPOSED RUL 7/ .

ACTION: Proposed rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is considering revising its regulations for the safe transportation of radioactive material to make them compatible with those of the International Atomic Energy Agency (IAEA)andthuswiththoseofmostmajornuclearnationsoftheworld.

Although several substantive changes are proposed in order to provide a l more uniform degree of safety for various types of shipments, the Comis-sion's basic standards for packaging radioactive material remain unchanged.

These regulations apply to all NRC specific licensees who place in transit byproduct, source, or special nuclear material. The Department of Trans-portation is also proposing a corresponding rule change to its Hazardous Materials Transport Regulations. In addition, three Petitions for Rule-making concerning the transportation of low-specific-activity (LSA) radio-active material are considered in this notice, and the criteria for approval of packages for the air transport of plutonium are proposed to be included in 10 CFR Part 71.

1 OATES: Submit coments by /d/s N , coments received after this date will be considered if it is practical to do so, but assurance 8810190384 000523 g t/NhV

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I I of consideration cannot be given except as to comments received on or j

! before this date. I 1

l ADDRESSES: Send comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555, ATTN: Docketing and Service Branch. '

Hand deliver comments to Room 11555 Rockville Pike, Rockville, MD., l l

4 between 7:30 a,.m,. and 4:15 p.m.

j Examine comments received and regulatory analysis at: The NRC ,

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Public Document Room, 1717 H Street NW., Washington, DC. j

) Obtain single copies of the regulatory analysis from: Donald R.

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1 Hopkins, Radiation Protection and Health Effects Branch, Office of Nuclear

! Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone: (301)492-3784, t j l FOR FURTHER INFORMATION CONTACT: Donald R. Hopkins, Radiation Protection f

and Health Effects Branch, Office of Nuclear Regulatory Research, U.S.

.l Nuclear Regulatory Commission, Washington, DC 20555, telephone: l (301)492-3784.

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SUPPLEMENTARY INFORMATION:

l j Background i On August 5,1983, the NRC published in the Federal Register (48 FR 35600) a final revision of 10 CFR Part 71, "Packaging and Transportation l of Radioactive Material." That revision, in combination with a paralle) 2

! revision of the hazardous materials transportation regulations of the

! U.S. Department of. Transportation (00T), brought United States domestic 4

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(7590-01) s transport safety regulations at the Federal level in accord with relevant j portions of the International Atomic Energy Agency (IAEA) design and per-i formance requirements to the extent considered feasible. This action made

! U.S. regulations compatible with the domestic regulations of most of the i

l. international community.

l In 1980, 1982, and November 1983, the IAEA assembled revision panels  !

t j to draft changes for the scheduled 1984 revision of its transportation regulations. The revision was eventually issued in sarly 1985. The l t

revisionpanels,whichconsistedofrepresentativesofmostmajorcountries

! involved in nuclear material transportation, made IAEA regulations more  ;

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) compatible with U.S. regulations theough some of their decisions. When  ;

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hRC and DOT were finalizing their transportation regulations in 1983, they  !

l anticipated some of the changes that IAEA was making in the revision of f l its regulations. The 1983 NRC and DOT rules were written to incorporate,  !

4 r I to the extent possible, some of the IAEA changes. Where it was not pos- I I

i j sible to incorporate IAEA changes, the 1983 rules were written so as to j j minimize the number of changes that would have to be made when the IAEA  !

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} revision was completed. Those changes .and other changes, not anticipated 1

by NRC and 007 in 1983, are being incorporated into this proposed -

] 7 l rulemaking, i

! f DiscussionofMajorChanges

The major proposed changes to 10 CFR Part 71 are additional accident 1 test requirements for certain packages, an expansion in the number of  !

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i radionuclides with listed limits for the quantity of radioactive material i

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! in a single package, a number of changes in the listed limits, simplifica- i

tion of the fissile material transport classes, updating requirements-i for shipment of low-specific-activity materials and inclusion of the criteria for air transport of plutonium. Thesemajorchangesaredis- f t

i cussed in the following paragraphs. d I  !

\ i Additional Accident Test Requirements l

! A deep water immersion test has been added to the regulations for a i Type B package contair,ing irradiated nuclear fuel in excess of 10'Ci  !

(37PBq). If such a package were lost in relatively shallow cotital f r

waters due tu the sinking or capsizing of a ship or barge, the probabil-l ity is high that an attempt would be made to recover the package and its [

6 j contents. The deep immersion test (200 m) requirement, which can be j l satisfied through engineering evaluation or actual physical test lr 6

j ($ 71.41), is to assure that the package containment system does not f

{ rupture from the water pressure at 200 m (656 ft) which would create j

radiological problems for the recovery operation or an additional environmental risk, i

i While the NRC staff believes that many existing Type 8 cask designs now approved by NRC will satisfy this additional test without need for modification, adding the test to the regulaticas will assure that foreign j casks and future U,5, designs will also havs the ability to survive deep  !

impersion in water, Adding the deep immersion test to the regulations also introduces the cost involved in having licensees analyze existing package designs to assure and demonstrate that presently approved casks l for transporting irradiated nuclear fuel satisfy the requirement. These costs can be avoided if no additional casks of the same design will be i  !

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I fabricated beyond a sper,f!e." ut. and if the casks will not be used for .

international transport m ia ,1cified date. In that case, existing casks of the approved deti<.i , Sais continue to be used domestically with no j further qualification regarding deep immersion. ,

r i A dynamic crush test has been added to the Type B package rules in i l

i addi;, ion to the 9 m (30 ft) drop test for packages which are minimally i vulnerable to damage in the drop test, but which have a high potential for radiation hazard if package failure occurs. The crush test requirement, I which can also be satisfied through provisions of $ 71.41, is applied to f packages which are both lightweight, up to 500 kg (1100 lb) and low density [

1 i (up to 1,000 kg/m3 , i.e., 3 g/cm ), 3 and which have a high radioactive l

material content (over 1000 A2 ) in normal form. The dynamic crush test I L

] consists of the drop of a 500 kg (1,100 lb) mild steel plate from 9 m onto  !

L the package resting on an unyielding support. IAEA applies the crush test  !

I in place of the 9 m drop test for the lightweight packages specified. In  ;

j the absence of experience using the crush test, and because the crush test and drop test evaluate different features of a package, NRC is requiring  !

i both the crush test and the 9 m drop test for lightweight packages. l l There are a limited number of light weight, low density, and high  !

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, radioactive material content package ,,. signs to which the crush test would  !

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l apply. Of those, some would pass the crush test so that no package design

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modification would be necessary. A limited amour.t of analysis would suffice to requalify the package design to the new standards. If the package destgr.

is not used internationally, and no further packages will be fabricated after a specific date, no effort to satisfy the new crush test standard i

would be necessary, and existing packages could be used in domestic '

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transportation to the end of their useftel lives.

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Changes in Radionuclide Limits l l The preamble to the August 5, 1983 revision of 10 CFR Part 71  !

I (48 FR 35600) noted that the IAEA, as part of its effort to maintain the continued adequacy of the regulations, had adopted a modified system for i determining As and As values. The .t.: and A2 values are the maximum quan-

) tity of a particular radionuclide permitted in Type A packages in speciai l form and normal form, respectively. Type A packages are those which pro- j l vide adequate containment, shielding, and criticality control under normal l I conditions of transport and minor accidents, but are not designed to sur-l vive severe transportation accidents. Instead, there are limits placed [

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on Type A package radioactive material contents. Accident resistant pack-ages are identified as Type B. Radioactive material in special form is .

] i either a nondispersible solid or sealed in a capsule so that the disper-l sibility, and therefore the radiological hazard, of the radioactive mate-1 rial is diminished. This systet ;f limiting the radioactive content of .

i Type A packages to At and A2 values depending on the dispersibility of 1

i the contents is the regulatory scheme for limiting the potential radio-i  !

logical hazard of a serious transportation accident involving packages of

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l radioactive material.

l The IAEA's modified system for determining As and A2 values is based on achieving essentially the same limitations on potential accident l

! radiological hazards as its predecessor system. However, the new system

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has the following advantages:

1. It states the radiation protection criteria employed more i

clearly;

2. It incorporates the data and conclusions on metabolic pathways l provided over the years 1977-1981 by the International Commission on (

l Radiological Protection (ICRP); '

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3. It includes dosimetric routes not previously considered; and  !

4. It harmonizes IAEA regulations with ICRP recommendations on radiological safety in Publications ICRP-26 and ICRP-30.1 i The effect of IAEA's adoption of this new system for calculating As a .

! and As values, and the subsequent incorporation of the new values in U.S. l domestic regulations, is that most current As and As values are changing

! in this revision. Of the 284 radionuelide entries in 10 CFR Part 71, A, i 1  ;

! values are being raised in 129 cases and lowered in 95 cases. Of the A [

values, 144 are being raised and 73 are being lowered. Based on our most current knowledge of radioactive material shipments in the United States 2 the economic impacts of these changes are not likely to be large. However,  ;

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any situations where a pntential exists for significant economic impacts f as a result of changes in the As or Aq values should be brought to the l l

Commission's attention in public comments, i l The new IAEA system for calculating As and A values is described l

)' in Appendix I, "The Q System for the Calculation of As and A Values," of  !

I IAEA Safety Series No. 7. "Explanatory Material for the IAEA Regulations l for the Safe Transport of Radioactive Material (1985 Edition)." Single  :

I r i copies of Appendix ! are available free of charge from the contact for  :

i 1 this rulemaking. I i i l'  !

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'ICRP publications are available for sale at Per anon Press, Inc.,  !

} Maxwell House, Fairview Park, Elasford, NY 1052 .  !

2 Transport of Radioactive Material in the United States, SRI Inter-  !

I national, SAND 84 7174 April 1985, is available from the National Tech- i j nical Information Service, U.S. Department of Commerce, 5285 Port l Royal Road, Springfield, VA 22161. [

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. Expansion of Radionuclide List I Based on numerous preposals for additions to the table of radio-  !

i nuclides in which limits are listed for the quantity of radioactive mate-

rial in a single package, IAEA concluded that its table needed to include

all radionuclides which have the potential for transportation. As a result,  !

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Table A-1 in Part 71, which provides Ag and A2 values, has been expanded  ;

from 284 entries to 378 entries. Because there now should be few instances l

where unlisted radionuclides would be transported, the rules for calculat- l 1 ing values for unlisted radionuclides have been simplified. The deter-  !

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4 mination of limits for unlisted radionuclides, oxcept for very conserva-  !

! i I tive values, will be made subject to Commission approval, j J

l i Simplification of Fissile Material Classes f

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! As a result of the evolution of the fissile material criteria, IAEA l j recognized that the current three fissile classes could be combined and  !

simplified into a single system. The effect of the simplification of i

) the IAEA system now being proposed for U.S. regulations is:  !

1. Elimination of the three fissile class designations;

2. Establishment of a single set of criteria for all packages of

! fissile materials;  !

l 3. Use of the transport index as the primary control of accumula-l tions of packages in transport under nearly all conditions; and l

4. Use of special arrangements for packages which do not meet the (

criteria.

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Updatino of Requirements for Shipment of LSA Material Over the last two major revisions of its transport regulations, IAEA has been working to update its requirements for shipment of LSA material to recognize the developing need for transportation of irradiated and con-taminated parts and equipment from decommissioned nuclear p; ants. Although I these developing LSA requirements were not factored into U.S. regulations when last updated in 1983, it is believed that the IAEA standards are now l mature enough to be adopted as U.S. standards.

Updating of the LSA regulations consists of the following:  :

1. An expansion of the LSA definition to include n6w types of material; l

2. Anewdefinitionof"surfacecontaminatedobject"(SCO)which  !

is treated in a manner similar to LSA material; and  !

3. An increase of specific activity limits for nondispersible, nonrespirable forms of LSA material while at the sama time limiting the l quantity of LSA material which can be shipped in other than a Type B  !

package. The package quantity limit is intended to limit external radia-tion levels produced as a result of shieldirg loss in a transportation i accident. l l

The NRC and 00T have overlapping statutory authority for the regula-tion of the transportation of radioactive material, so the regulations .f either or both agencies may apply, depending on the circumstances involved. l l

In order that DOT may act as the only regulator of LSA materials and SCO in l quantities b low those where external radiation levels become important, the NRC is proposing a new exemption in i 71.10. This provision would exempt licensees from most provisions of 10 CFR Part 71 for shipment and ,

I carriage of LSA/SCO materials which can be transported in bulk without l 9

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h par' ages (LSA-I and SCO-I), and also in their shipment and carriage of LSA/SCO materials in packages containing up to a 2Ai quantity of radio-active materiai. At this level of activity, NRC regulations become applicable and Type B packages, the design of which must be approved by NRC, are required. This action, if adopted, would have the effect of raising the threshold level of radioactivity at which NRC regulates ship-ments of LSA radioactive material from the Typo B quantity level to the level at which Type B packages are required.

Although the regulations proposed by the NRC at S 71.10 specify 2Ai l as the level of coauined radioactive material which causes NRC regula-tions and Type B packeges to become applicable, the IAEA corresponding standard is expressed as a') external radiation level at 3 m from the 1

unshielded material or object of I rem /hr (10 mSv/h). The value A for i

l any specified radionuclide is the quantity of that radionuclide as a point source which produces a radiation level of I rem /h at a distance of 3 m.

Considering that LSA and SCO materials are bulk sources with considerable I self-shielding, the value 2Ai was chosen as a close approximation of the IAEA standard of 1 rem /h at 3 m. The NRC staff, in implementing this IAEA standard in 10 CFR Part 71, takes the posftion that the radiation level standard would be very difficult for the industry to apply, and that g expressing the limit in units of radioactivity would be a more reasonable approach. The approach recommended by the staff will make U.S. reguia-1 tions inconsistent witn those of IAEA, although in an area where there is little international transport. The NRC is particularly interested in whether industry shares the NRC staff view that this limit should be expressed as a limit on radioactivity because a radiation level limit as adopted by IAEA is impractical for the industry to implement.

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o Part 71 Inclusion of Criteria for Air Shipment of Plutonium As a result of Congressional action in 1975, Public Law 94-79 prohibited the NRC from licensing the air shipment of plutonium in any form until the NRC certified to the Congress that a safe container had been developed and tested such that the container "will not rupture under crash and blast-testing equivalent to the crash and explosion of a high-flying aircraft." The NRC developed and certified to the Congress package criteria which it believed corresponded to the Public Law and published these criteria in NUREG-0360, "Qualification Criteria to Certify a Package for Air Transport of Plutonium," dated January 1978. This rule-making action would amend 10 CFR Part 71 to include these criteria in SS 71.64, 71.74, and 71.88.

It is the Commission's view that the import / export or domestic transport of plutonium by air pursuant to the reqiirements of Pub.

L. 94-79, as implemented by SS 71.64, 71.74, and 71.88 of this Part, is not affected by Section 5062 of Pub. L. 100-203. Certification of containers for the air transport of plutonium for shipments subject to Section 5062 and the development of appropriate test criteria for such certification are not within the scope of this proposed rule. Thuse matters will be considered by the Commission separately from this rulemaking action.

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o Detailed Changes Detailed substantive changes as proposed by the NRC for public comment are described in the following paragraphs, arranged by section number:

1. Section 71.4, "Definitions," would be amended as follows:

The definition of fissile classes would be deleted to correspond to the major change of eliminating fissile classes. Fissile material would be defined as the listed radionuclides, and the definition of fissile radionuclides would be deleted.

The definition of low-specific-activity (LSA) material would be extensively changed to correspond to that of IAEA. The one remaining significant difference would be the addition of a provision in NRC regula-tions for transportation of contaminated earth in a closed vehicle in unpackaged form. Extensive removal of contaminated earth has been found necessary in decommissioning facilities in the United States, a process apparently not yet required in most other IAEA Member States. Most LSA material woula be subsequently exempted from Part 71 control by the pro-visions of Section 71.10, "Exemption for low level materials." The 00T regulations would specify the requirements for packaging LSA material.

The grandfather clause for special form radioactive material encapsulation would be updated.

A new definition of Surface Contaminated Object (SCO) would be added to correspond to the parallel definition in IAEA regulations.

SCO would be treated in the regulations .similarly to LSA materials, with industrial packaging required for most applications. As with LSA 12 1

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materials, most SCO would be exempted from Part 71 control by the provi-sions of Section 71.10, "Exemption for low level materials."

Some progress has been made in expressing radiological limits in dual units, in that limits currently expressed in units of Rems and Curies now also show the International System of Units (SI) equi-valents in Sieverts and Bequerels along with the customary units. In most cases the limits in customary units have been extended to 3 significant figures so they are equal to the limits expressed in SI units to a tenth of a percent. Limits on length, pressure, weight, and termperature are

- expressed in SI units in the current 10 CFR Part 71, with approximate values in customary units following in parentheses. Those values in customary units have been extended to 3 significant figures to make them equal to the limits expressed in SI units. The objective of this approach is to maintain' consistency with international regulations while allowing U.S. shippers to use the units with which they are most familiar. In the case of the special limite, on shipmentt of plutonium in NRC regulations, for which there are no comparable international rules, the limits expressed in SI units have been carried out to three significant figures to make them i equal to the U.S. limits expressed in customary units. The end result of this effort is that licensees can use either limit expressed in the regulations as they are considered to be equal. The Commission is parti-cularly interested in public comments on this method of expressing dual units in the regulations.

2. Section 71.5, "Transportation of licensed material," would be amended to correct a number of referencing errors.

3. Section 71.10 "Exemption for low level materials," would be amended to include exemptions for LSA material and SCO. The categories 13

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LSA-I and SCO-I would be limited to very low concentrations of radio-active material which would be allowed to be transported unpackaged.

TMir exemption would be specified separately in a new paragraph 71.10(c),

without restriction on total quantity. LSA-II and LSA-III materials and SCO-II would be required to be packaged and thus would be specified in paragraph 71.10(b) of the regulations with a package quantity limit as explained earlier in this preamble.

4. Section 71.12, "General License: NRC-approved package," would be amended in paragraph (e) to clarify that previously approved fissile material packages would be subject to the restrictions of S 71.13, "Previously approved package."

5. Section 71.13, "Previously approved package," would be amended to update the restrictions for packages aoproved under previous editions of the regulations. In line with parallel provisions of IAEA transport regulations, packages acceptable under the 1967 NRC transport regulations (which correspond to the 1967 IAEA regulations) can no longer be manufac-tured for use but may continue to be used. These packages must be marked with a unique serial number for identification and control. International use of these packages requires multilateral approval of all countries involved in their tie. Packages acceptable under the 1983 NRC transport regulations (which correspond to the 1973 IAEA regulations) can be manufactured until the end of 1995. They will be subject to multilateral approval for international use of the package after 1992. Approvals for any package design can be upgraded to present status through an applica-l tion which demonstrates that current standards are satisfied. l

6. Section 71.14 "General license: 00T specification container,"

would be amended to reflect the 1985 IAEA grandfathering provisions.

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7. Sections 71.16 - 71.24, general licenses for foreign approved packages and fer fissile material packages, would be amended to clarify that those general licenses are subject to the quality assurance require-ments of Subpart H of Part 71, requirements already imposed by S 71.101, but whose applicability has been misunderstood by some persons. Minor technicol changes have also been introduced to make those general license provisions correspond to standards in IAEA transport rules.

8. Section 71.31, "Contents of application," would be amended so that S 71.31(a)(3) may be satisfied by submittal of a quality d assurance program description" as required by S 71.37 or by "reference to a pre-viously approved quality assurance program" in an application for package design approval. Whether or not an applicant has a previously approved quality assurance program to which it can refer, the applicant should recognize that the package design work necessary to develop the descrip-tions included in its application for NRC approval must be done under the quality assurance program eventually approved by NRC regarding that pack-age design. To avoid the situation where package design work is invali-dated because changes become necessary in a quality assurance program under which the package design work was done, an applicant may wish to obtain approval of its quality assurance program prior to investing a large amount of effort in the package design program. The NRC encourages new applicants, who do not yet have NRC approved quality assurance programs, to obtain at least a partial approval of the design portion of the program.

In addition, S 71.37(b) was moved to S 71.31(c) as a more appropriate location.

9. A new section 71.38, "Renewal of a certificate of compliance or quality assurance program approval," would extend the concept of "timely l

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renewal" from the NRC licensing regulations in 10 CFR Parts 30, 40, and 70 to the package design and quality assurance approvals in 10 CFR Part 71.

Submittal of an application for renewal of a package certificate of com-pliance or quality assurance approval at least 30 days prior to its expiration would automatically extend the expiration date of the existing approval until the NRC makes a final decision regarding the application.

The provision also would require that a renewal application consolidate the prior approval and all subsequent revisions.

10. Section 71.43, "General standards for all packages," would be

. amended as follows:

Paragraph (c) would require that a positive fastening device protect against a rise in internal pressure; Paragraph (d) would require that behavior of materials under irradiation be considered in assuring the absence of significant chemical, galvanic, or other reaction among package components and the package contents; Paragraph (f) would continue to require that there be "no significant increase" in external radiation levels as a result of sub-jecting a package to the normal conditions of transport. The IAEA has quantified that increase as being no more than 20 percent, a value the NRC staff believes is too large for the consequences of normal handling or minor accidents which can occur more than once during transport and for which no corrective action normally would be taken. The NRC proposes to continue to approve shielding retention of package designs on an ad hoc basis until what it considers to be a more reasonable standard is determined.

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Paragraph (h) would continue to prohibit continuous venting during transport but would allow intermittent venting when the associated operational controls are approved by the Commission. While the IAEA regulations no longer prohibit continuous venting, that prohibition would be continued in NRC transport regulations because the staff considers continuous venting to be poor engineering practice.

11. Section 71.51, "Additional requirements for Type B packages,"

would be amended as follows:

Paragraph (a)(1), as in S 71.43(f), would continue to require that there be "no significant increase" in external radiation levels as a result of subjecting a package to the normal conditions of transport not-withstanding the IAEA's determination that a 20 percent increase con-stitutes no significant increase; and Paragraph (a)(2) would reduce allowable krypton-85 releases under the hypothetical accident conditions from 10,000 Ci to 10 A2 or 2,700 Ci (10TBq).

12. Section 71.52, "Exemption for low-specific-activity (LSA) packages," would be written as two exemptions from certain Type B pack-age requirements for packages containing only LSA material and trans-ported as exclusive use. The broader of these exemptions, in paragraph (b), would be identical to the present exemption, but would be condl-tioned to expire one year after the effective date of these amendments.

While it is in effect, the broader exemption of paragraph (b) would deal with nonapplicability of accident resistance requirements to quantities i of LSA material in excess of Type B quantities in a single package, recognizing the very low toxicity of low-specific-activity radioactive material. Elimination of that provision would subject LSA material in l 17 I

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excess of the quantity 2At in a single package to all Type B package standards including the hypothetical accident conditions. The one year delay in discontinuing this exemption is intended to allow the industry to develop and fabricate enough Type B waste packages to satisfy the need which would be brought on if this change were adopted. Information pro-vided to NRC indicates that only 5 Type B waste packages are now in existence, while many more would be needed to satisfy the need which would be created if this proposed change were adopted.

The more narrow exemption of paragraph (a) would continue to recognize the low toxicity of LSA material, but to a lesser extent than paragraph (b), by providing an exemption from the Type B requirement in S 71.51(a)(1) which limits the loss or dispersal of radioactive contents under normal conditions of transport. That provision requires that leak-tightness of a' Type B package be demonstrated to a sensitivity of 10 8 A2 , a specification unnecessary for the low toxicity LSA material.

Although it would be exempt from the sensitivity provision, the LSA pack-age design would still have to satisfy the general standard of S 71.43(f) that there be no loss or dispersal of radioactive contents as a result of subjecting a package to the normal conditions of transport.

13. Section 71.53, "Fissile material exemptions" would be amended as follows:

Present S 71.53(b), specifying an exemption for natural and depleted uranium which has been irradiated in a thermal reactor, would be deleted because the material described would, by definition, no longer be fissile material; and I

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The existing paragraph (g) would be redesignated S 71.53(f),

and it would place an additional limitation on the nitrogen-to-uranium atomic ratio.

14. Section 71.57, "Specific standards for a Fissile Class I package," and Section 71.61, "Specific standards of a Fissile Class III shipment," would be deleted because the three fissile classes would be combined into new section 71.59, "Standards for arrays of fissile material packages."

15. A new section 71.61, "Special requirement for irradiated nuclear fuel shipments," imposing a deep water immersion test would be added. ,

16. Section 71.63, "Special requirements for plutonium shipments,"

would be revised to accept a suggestion received from the E. I. DuPont Savannah River Plant during the last major revision of 10 CFR Part 71.

The suggestion was that the special requirements for solid form and double containment now applied to shipments of all isotopes of plutonium be applied only to the extremely radiotoxic isotopes of plutonium (excluding plutonium-241) and to other extremely radiotoxic radionuclides as well (including, for example, americium-241 and actinium-227). While this suggestion was favorably received, it was beyond the scope of that rule-making action and is proposed now. While the change seems reasonable from the health and safety standpoint, any significant technical and ,

economic impacts of the change should be included in comments to the Commission so they may be considered.

17. Section 71.73, "Hypothetical accident conditions," would be amended to add a dynamic crush test for certain packages, and to make minor modifications to the thermal test in accordance with changes e de 19

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to IAEA regulations. Times specified for the immersion tests seem superfluous and have been deleted. Determination of acceptance under the standard should not depend on the time of immersion.

18. Section 71.75, "Qualifications of special form radioactive materials," would add an alternative method to qualify a special form capsule under the tests prescribed in the specified standard of the International Organization for Standardization (ISO).

19. Section 71.77, "Tests for special form radioactive material,"

would add an alternative method to qualify special form radioactive material under the specific impact and temperature tests prescribed in the specified standard of the 150.

20. Section 71.95, "Reports," would include a new paragraph (c) to require reporting by a licensee if any conditions of approval in the certificate of approval were not oberved in making a shipment.

21. Section 71.97, "Advance notification of shipment of nuclear waste," would be amended to redefine the level of radioactivity at which advance notification is required for shipments of spent nuclear fuel and radioactive waste to make that level more uniform across the range of radionuclides transported. The new level specified would correspond to that at which the DOT imposes its routing and training requirements, and to that at which IAEA imposes additional administrative requirements such as multilateral shipment approvals. The effect of this change is expected to decrease the overall number of packages subject to advance notification i and to increase reporting of packages containing large amounts of tran-suranic alpha-emitting nuclides.

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Other Regulatory Actions Three petitions for rulemaking were filed with the NRC in connection with the rules for transporting LSA radioactive material. The substance of the three petitions was essentially the same, to request that NRC exempt LSA materials from its requirements in 10 CFR Part 71. This would have left the regulation of all LSA material to the 00T. The control of LSA material, as with the control of all radioactive material, was divided at that time, as it today, between NRC and 00T. 00T controlled carriers and shippers of small quantities of all radioactive materials through provi-sions in its regulations in 49 CFR, while NPC controlled shippers of fissile material and of larger quantities of other radioactive materials through its regulations in 10 CFR and its licensing program.

The petitioners were the Energy Research and Development Administra-tion (now the U.S. Department of Energy) in its letter dated July 23, 1975 (PRM-71-1); the American National Standards Institute (ANSI) Committee N14 in its letter dated March 10, 1976 (PRM-71-2); and Chem-Nuclear Systems, Inc., in its letter dated November 22, 1976 (PRM-71-4). All three petitioners argued that the control NRC was exerting over transportation of LSA materials created an inconsistency between NRC regulations and those of the IAEA and should be discontinued. A proposed rule that would  !

have provided the exemption for LSA materials requested in the petitions was published by NRC for public comment on August 17, 1979 (44 FR 48234).

Prior to finalization of that rule, however, a deficiency in the new LSA requ' 's, as proposed, was recognized so that the entire LSA proposal, includi.3  : exemption, was withdrawn. In the interim, the deficiency I 21 1

(7590-01]

in the LSA requirements in the IAEA regulations was recognized and corrected. That correction is discussed under the "major changes" section of this preamble. The correction introduces a distinction between the requirements for small quantities of LSA material and those for larger quantities. This distinction is implemented in the U.S. regulatory scheme as one set of requirements in D0T regulations for small quantities of LSA material and as a different set of requirements in NRC regulations for larger quantities of LSA material.

As a result of these changes in LSA requirements, the exemption requested in the three petitions cannot be provided. The requirements proposed for inclusion in NRC regulations are consistent with the regula-tory schemes of both DOT and IAEA. Because the level of radioactivity at which NRC controls are imposed in the proposed rule is somewhat higher than in the current rule, there is an exemption provided in S 71.10 for LSA materials up to the level where NRC regulations impose additional packaging requirements. This exemption is of limited scope, however, and does not satisfy the intent of the petitions. For the above reasons, the NRC plans to deny the three petitions if changes proposed for the LSA por-tions of this rulemaking are carried forward to the final rule.

Finding Of No Significant Environmental Impact: Availability The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment and therefore an environmental impact statement is not required.

22

[7590-01)

The Commission's "Final Environmental Statement on the Transportation of Radioactive Material by Air and Other Modes," NUREG-0170,3 dated December 1977, is NRC's generic environmental impact statement (EIS) covering all types of radioactive material transportation by all modes (road, rail, air, and water). From the Commission's latest survey of radioactive material shipments and their characteristics, "Transport of Radioactive Material in the United States," SAND 84-7174, April 1985, we can conclude that current radioactive material shipments are not so different from those evaluated in NUREG-0170 as to invalidate the results or conclusions of that EIS. Environmental impacts associated with this 1

proposed rulemaking are evaluated in "Regulatory Analysis of Proposed Changes to NRC Regulations on Packaging and Transportation of Radioactive Material," draft dated November 1987, prepared for NRC by Pacific North-west Laboratory, Richland, Washington.

NUREG-0170 established the non-accident related radiation exposures associated with transportation of radioactive material in the United States as 9800 person-rem which, based on the conservative linear radia-tion dose hypothesis, resulted in 1.7 genetic effects and 1.2 latent cancer effects per year. More than half this impact resulted from shipment of medical-use radioactive materials. Accident related impacts were estab- ,

lished at about 1 genetic effect and 1 latent cancer fatality for 200 years of transporting radioactive materials. The principal nonradiological impacts were found to be 2 injuries per year and less than 1 accidental death per 4 years. In contrast, non-accident related radiation exposures 3 Copies of NUREG-0170 may be purchased tnrough the U.S. Government Print-ing Office by calling (202)275-2060 or by writing to the U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-7082.

23 1

I I l

y, associated with this rulemaking would be increased by 2 person-rem /y

(.02 person-sieverts/y) while accident related impacts would be decreased by 4.5 person-rem /y (.045 person-sieverts/y). Nonradiological traffic injuries would be increased by 0.24 per year and by 0.012 traffic deaths per year (less than 1 accidental death per 80 years). These impacts are judged to be insignificad ..ompared to the baseline impacts established in NUREG-0170.

The environmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Document Room, 1717 H Street NW, Washington, DC. Single copies of l the environmental assessment and finding of no significant impact on which l

l this determination is based are available for inspection at the !!RC Public Document Room, 1717 H Street NW, Washington, DC. Single copies of the environmental assessment anct finding of no significant impact are also available from Donald R. Hopkins, Radiation Protection and Health Effects Branch, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, Telephone: (301) 492-3784.

Paperwork Reduction Act Statement The proposed rule would amend information collection requirements that are subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et seq.).

This rule has been submitted to the Office of Management and Budget for review and approval of the paperwork requirements.

24

[7590-01]

Regulatory Analysis The Commission has prepared a draft regulatory analysis on this pro-posed regulation. The analysis examines the costs and benefits of the alternatives considered by the Commission. The draft analysis is avail-able for inspectior, in the NRC Public Document Room,1717 H Street NW. ,

Washington, DC. Single copies of the analysis may be obtained from the contact identified earlier.

The Commission requests public comment on the draft analysis. Com-ments on the draft analysis may be submitted to the NRC as indicated under i

the ADDRESSES heading.

Backfit Analysis The factors which must be considered in a backfit analysis associ-ated with changes in transportation regulations are evaluated in the i

"Regulatory Analysis of Proposed Changes to NRC Regulations on Packaging and Transportation of Radioactive Material," draft dated November 1987.

That evaluation shows very small changes in accident risks as a result of q the adoption of the proposed revision, but some reduction in maximum consequences given an accident. The evaluation shows broad improvement in NRC regulatory consistency with IAEA at an initial cost of $1,800,000 ,

and a continuing annual cost of $1,700,000 per year (Table S.1).

1 The continuing costs are associated with the addition of a new limit on the quantity of LSA radioactive material allowed in a single transpor-

! i tation package. This new limit is considered internationally to be a l I

necessary safety requirement to limit the consequences of a severe i

transportation accident involving LSA material. f 4

25

[7590-01)

The initial costs are chiefly associated with industry upgrading of its package safety analyses to include the proposed new accident crush and immersion tests and with the NRC review of those new analyses. The estimated costs are overstated because of the assumption that all licen-sees using packages approved under earlier regulatory standards would take immediate steps to upgrade the package analyses so the package approvals would reflect approval under the latest revised standards.

, Although that is a prudent assumption absent any reasonable basis for predicting actual licensee reaction, there is little reason licensees would take any immediate action to upgrade their package approvals. Both domestic and international regulations are based on the responsible agency's confidence that packages built to n design approved under earlier standards are adequately safe for continued use, although new package construction to that design would be limited and international use after 1992 would requira approval by all countries through which the package is to be transported. In actual practice, some package approvals would ,

never be upgraded; those that would be upgraded would be done over a period of several years as guidance and experience in upgrading becomes available. ,

1 Although the regulatory analysis shows a small reduction in accident )

i risks from the proposed changes and some reduction in maximum conse-quences given an accident, the primary benefit of this rulemakir.g action j would be to achieve consistency in radioactive material transportation regulations between the United States and the rest of the world. This consistency would not only facilitate the free movement of radioactive u.aterials between countries for medical, research, industrial, and nuclear fuel cycle purposes, but it would also contribute to safety by l

26 l

- -, . , - . - _.-,-__.----c - , _ - - _ _ _ _ , . - - - m ,_ - - - - - - - -

/-

[7590-01]

concentrating the efforts of the world's experts on a single set of safety standards and guidance (those of the IAEA) from which individual countries could develop their domestic regulations. In addition, the accident experience of every country that bases its domestic regulations on those of the IAEA could be applied to every other country with consis-tent regulations to improve its safety program.

In summary, the benefits and costs associated with this proposed rulemaking are of two categories. The first consists of changes to make U.S. regulations compatible with those of the IAEA. This effort provides major benefits including a substantial increase in the overall protection of the public health and safety, and it is associated with short-term and relatively minor costs which are justified in view of this increased pro-tection. The second category consists of the additional limit imposed on shipments of LSA material. This effort is associattd with significant ongoing costs, but internationally the new limit is considered to be a necessary safety requirement to limit the consequences of a severe transportation accident involving LSA material.

Regulatory Flexibility Act Certification In accordance with the Regulatory Flexibility Act of 1980, (5 U.S.C. 605(b)), the Commission certifies that this rule will not, if promulgated, have a significant economic impact on a substantial number of small entities. This proposed rule affects NRC licensees, including operators of nuclear power plants, who transport or deliver to a carrier for transport relatively large quantities of radioactive material in a single package. These companies do not generally fall within the scope 27

/-

. [7590-01]

of the definition of "small entities" set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regula-tions issued by the Small Business Administration at 13 CFR Part 121.

List of Subjects in 10 CFR Part 71 Hazardous materials transportation, Incorporation by reference, Nuclear materials, Packaoing and containers, Penalty, Reporting and recordkeeping requirements.

t For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 553, the NRC is considering adoption of the following amendments to 10 CFR Part 71.

PART 71 - PACKAGING AND TRANSPORTATION OF RADI0 ACTIVE MATERIAL Subpart A - General Provisions Ses.

7

1.0 Purpose and Scope

71.1 Communications.

71.2 Interpretations.

71.3 Requirement for license.

1 71.4 Definitions. '

71.5 Transportation of licensed material.

Subpart B - Exemptions 71.6 (Reserved.]

71.7 Specific exemptions.

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l 71.8 [ Reserved.]

71.9 Exemption of physicians.

71.10 Exemption for low-level materials.

71.11 (Reserved.]

Subpart C - General Licenses 71.12 General license: NRC-approved package.

71.13 Previously approved package.

71.14 General license: 00T specification container.

' 71.16 General license: Use of foreign approved package.

71.18 General license: Fissile material, limited quantity per package.

71.20 General license: Fissile material, limited moderator per package.

71.22 General license: Fissile material, limited quantity, controlled shipment.

71.24 General license: Fissile material, limited moderator, controlled shipment.

Subpart 0 - Application for Package Approval 71.31 Contents of application.

71.33 Package description.

71.35 Package evaluation.

71.37 Quality assurance.

71.38 Renewal of a certificate of compliance or quality assurance program approval.

71.39 Requirement for additional information.

29

m. - - -~

Subpart E - Package Approval Standards 71.41 Demonstration of compliance.

71.43 General standards for all packages.

71.45 Lifting and tie-down standards for all packages.

71.47 External radiation standards for all packages.

71.51 Additional requirements for Type B packages.

71.52 Exemption for low-specific-activity (LSA) packages.

71.53 Fissile material exemptions.

71.55 General requirements for fissile material packages.

71.57 Reserved 71.59 Standards for arrays of fissile material packages.

71.61 Special requirement for irradiated nuclear fuel shipments, j 71.63 Special requirements for plutonium and other high toxicity l radionuclide shipments.

71.64 Special requirements for plutonium air shipments.

71.65 Additional requirements.

Subpart F - Package and Special Form Tests 71.71 Normal conditions of transport.

~

71.73 Hypothetical accident conditions. i 71.74 Plutonium accident conditions.

71.75 Qualification of special form radioactive material. l l

71.77 Tests for special form radioactive material. '

Subpart G - Operating Controls and Procedures 71.81 Applicability of operating controls and procedures.

{

71.83 Assumptions as to unknown properties.

30

71.85 Preliainary d0 terminations.

[ 71.87 Routine determinations.

71.88 Air transport of plutonium.

71.89 Opening instructions.

71.91 Records.  ;

71.93 Inspection and tests.  ;

71.95 Reports. l 71.97 Advance notification of shipment of nuclaar waste.

71.99 Violations.  !

4 i i

i t

Subpart H - Quality Assurance t 71.101 Quality assurance requirements. [

71.103 Quality assurance organization.

71.105 Quality assurance program.  ;

j 71.107 Package design control.

71.109 Procurement document control. ,

71.111 Instructions, procedures, and drawings.

71.113 Document control.

j 71.115 Control of purchased material, equipment, and services. ,

71.117 Identification and control of materials, parts, and components.

71.119 Control of special processes.  ;

i 71.121 In9rnal inspection. f 71.123 Test control.

l 71.125 Control of measuring and test equipment. I

71.127 Handling, storage, and shipping control. l

71.129 Inspection, test, and operating status, i 71.131 Nonconforming materials, parts, or components.

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c svew ms2>

71.133 Corrective action.

71.135 Quality assurance records.

71.137 Audits.

Appendix A--Determination of At and A2 Authority: Secs. 53, 57, 62, 63, 81, 161, 182, 183, 68 Stat. 930, 932, 933, 935, 948, 953, 954, as amended (42 U.S.C. 2073, 2077, 2092, 2093, 2111, 2201, 2232, 2233); secs. 201, as amended, 202, 206, 89 Stat.

1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).

Section 71.97 also issued under sec. 301, Pub. L.96-295, 94 Stat.

789-790.

For the purposes of sec. 223, 68 Stat. 958, as amended (42 U.S.C.

2273), SS 71.3, 71.43, 71.45, 71.55, 71.63(a) and (b), 71.83, 71.85, 71.87, 71.89, and 71.97 are issued under sec. 161b, 68 Stat. 948, as amended (42 U.S.C. 2201(b)); and $$ 71.5(b), 71.91, 71.93, 71.95, and 71.101(a) are issued under sec. 161o, 68 Stat. 950, as amended (42 U.S.C.

2201(o)).

Subpart A - General Provisions S 71.0 Purpose and scope.

(a) This part establishes--

(1) Requirements for packaging, preparation for shipment, and transportation of licensed material; and (2) Procedures and standards for NRC approval of packaging and shipping procedures for fissile material and for a quantity of other licensed material in excess of a Type A quantity.

32

[7590-01] .

)

i (b) The packaging and transport of licensed material are also subject to other parts of this chapter (e.g., Parts 20, 21, 30, 40, 70, and 73) and to the regulations of other agencies (e.g., the U.S. Depart-ment of Transportation (DOT) and the U.S. Postal Service 1

) having juris-diction over means of transport. The requirements of this part are in addition to, and not in substitution for, other requirements.

(c) The regulations in this part apply to any licensee authorized j by specific license issued by the Commission to receive, possess, use, i or transfer licensed material if the licensee delivers that material to a carrier for transport or transports the material outside the confines of the licensee's facility, plant, or other authorized place of use. No provision of this part authorizes possession of licensed material.

(d) Exemptions from the requirement for license in S 71.3 are specified in S 71.10. General licenses for which no NRC package approval is required are issued in SS 71.14-71.24. The general license in S 71.12 requires that an NRC certificate of compliance or other package approval be issued for the package to be used under the general license. Applica-tion for package approval must be completed in accordance with Subpart D  !

of this part, demonstrating that the design of the package to be used satisfies the package approval standards contained in Subpart E of this part as related to the tests of Subpart F of this part. The transport of licensed material or delivery of licensed material to a carrier for trans-port is subject to the operating controls and procedures requirements of Subpart G of this part, to the quality assurance requirements of Subpart H 1 Postal Service Manual (Domestic Mail Manual), section 124.3, which is i incorporated by reference at 39 CFR 111.1 (1974).

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. . . [7590-013 of this part, and to the general provisions of Subpart A of this part, j including 00T regulations referenced in S 71.5.

S 71.1 Comunications.

All communications concerning the regulations in this part should {

be addressed to the Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555, or may be delivered in person at the Commission offices at 1717 H Street NW.,

Washington. 00, or its offices at 11555 Rockville Pike, Rockville, Maryland, i

1 L S 71.2 Interpretations.

]

Except as specifically authorized by the Commission in writing, no interpretation of the meaning of the regulations in this part by any officer or employee of the Commission other than a written interpreta-

[

tion by the General Counsel will be recognized to be binding upon the i Commission.  !

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5 71.3 Requirement for license.

Except as authorized in a general license or a specific license 4

issued by the Commission, or as exempted in this part, a licensee subject to the regulations in this part may not--

(a) Deliver any licensed material to a carrier for transport; or (b) Transport licensed material.

I 5 71.4 Definitions.

l The following terms are as defined here for the purpose of this part.

1 To ensure compatibility with international transportation standards, all 1

J 34

[7590-01]

limits in this part are given in terms of dual units: The International System of Units (SI) followed or preceded by U.S. standards or customary units. The U.S. customary units are not exact equivalents but are rounded to a convenient value providing a functionally equivalent unit.

For the purpose of this part, either unit may be used.

"Ai " means the maximum activity of special form radioactive material permitted in a Type A package. "A2 " means the maximum activity of radio- ,

active material, other than special form radioactive material, permitted in a Type A package. These values are either listed in Appendix A of this part, Table A-1, or may be derived in accordance with the procedure prescribed in Appendix A of this part, "Carrier" means a person engaged in the transportation of passengers or property by land or water as a common, contract, or private carrier, or by civil aircraft.

I "Close reflection by water" means immediate contact by water of sufficient thickness for maximum reflection of neutrons.

"Containment system" means the assembly of components of the pack-aging intended to retain the radioactive material during transport.

"Conveyance" means any tehicle, aircraft, vessel, or hold, compart-ment, or defined deck area of a vessel.

"Exclusive use" (also referred to in other regulations as "sole use" or "full load") means the sole use of a conveyance or of a large freight container with a minimum length of 6 m (19.7 ft) by a single consignor and for which all initial, intermediate, and final loading and unloading are carried out in accordance with the direction of the consignor or con-signee. Any loading or unloading must be performed by personnel having i

radiological training and resources appropriate for safe handling of 35 l

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[7590-05)~

the consignment. Specific instructions for maintenance of exclusive use shipment controls must be issued in writing and included with the shipping paper information provided to the carrier by the consignor.

"Fissile material" means plutonium-238, plutonium-239, plutonium-241, L uranium-233, uranium-235, or any combination of these radionuclides.

Unirradiated natural uranium and depleted uranium, and natural uranium or depleted uranium which has been irradiated in thermal reactors only are not included in this definition. Certain exclusions from fissile material  ;

l controls are provided in S 71.53.

"Low-Specific-Activity (LSA) material" means radioactive material ,

l with limited specific activity which satisfies the descriptions and i limits set forth below. Shielding materials surrounding the LSA mate-rial may not be considered in determining the estimated average specific l activity of the package contents. LSA material must be in one of three  !

]

groups:

i (1) LSA-I (i) Ores containing only naturally occurring radionuclides (e.g.,

uranium, thorium) and uranium or thorium concentrates of such ores; or (ii) Solid unirradiated natural uranium cr depleted uranium or l natural thorium or their solid or liquid compounds or mixtures; or  ;

(iii) Radioactive material, other than fissile material, for which l

the A2 value is unlimited; or  ;

(iv) Contaminated earth in a closed transport vehicle for which the estimated average specific activity does not exceed 10-8A /g. .

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[7590-01]

l (2) LSA-II (i) Water with tritium concentration up to 27.0Ci/2 (ITBq/2); or (ii) Material in which the radioactive material is distributed throughout and the estimated average specific activity does not exceed 10"A2 /g for solids and gases, and 10-5A 2 /g for liquids.

(3) LSA-III Solids (e.g. , consolidated wastes, actitated materiah) in which:

(i) The radioactive material is distributed throughout a solid or 4

a collection of solid objects, or is esser;ially unitormly distributed in a solid compact binding agent (such as concrete, bitumen, ceramic, etc.); and (ii) The radioactive material is relatively insoluble, or it is intrinsically contained in a relatively insoluble material, so that, even 4

under loss of packaging, the loss of radioactive material per package by .

i leaching when placed in water for seven days would not exceed 0.1 A2 ; and (iii) The estimated average specific activity of the solid does not exceed 2 x 10-3A /g.

"Maximum normal operating pressure" means the maximum gauge pressure that would develop in the containment system in a period of one year under the heat test specified in S 71.71(c)(1), in the absence of vent-f l J ing, external cooling by an ancillary system, or operational controls ,

dering transport. l l

"Natural thorium" means thorium with the naturally occurring l j

distribution of thorium isotopes (essentially 100 weight percent r

i thorium-232). ,

l I I

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,, (7590-01)

"Normal form radioactive material" means radioactive materi ' which l

has not been demonstrated to qualify as "special form radioactive material." l "Optimum interspersed hydrogenous moderation" means the presence of hydrogenous material between packages to such an extent that the maximum nuclear reactivity results.

"Package" means the packaging together with its radioactive contents as presented for transport. t (1) "Fissile material package" means a fissile material packaging together with its fissile material contents.

(2) "Type B package" means a Type B packaging together with its radioactive contents. On approval, a Type B package design is designated by NRC as B(U) unless the package has a maximum normal operating pressure of more than 700 kPa (102 lb/in2 ) gauge or a pressure relief device which would allow the release of radioactive material to the environment under the tests specified in S 71.73 (hypothetical accident conditions), in which case it will receive a designation B(M). B(U) refers to the need for unilateral approval of internationa.1 shipments; B(p.) refers to the need for multilateral approval of international snipments. There is no distinction made in how packages with these designations may be used in domestic transportation. To determine their distinction for interna-tional transportation, see 00T regulations in 49 CFR Part 173. A Type B package approved prior to September 6, 1983 was designsted cnly as Type B.

Limitations on its use are specified in S 71.13.

"Packaging" means the assembly of components necessary to ensure compliance with the packaging requirements of this part. It may consist of one or more receptacles, absorbent materials, spacing structures, 38

LiWu-ulj thermal insulation, radiation shielding, and devices for cooling or absorbing mechanical shocks. The vehicle, tie-down system, and auxiliary equipment may be designated as part of the packaging.

"Special form radioactive material" means radioactive material which satisfies the following conditions:

(1) It is either a single solid piece or is contained in a sealed capsule that can be opened only by destroying the capsule;

'2) The piece or capsule has at least one dimension not less than 5 mm (0.197 in); and (3) It satisfies the requirements of S 71.75.

A special form encapsulation designed in accordance with the requirements of 9 71.4 of this part in effect on June 30, 1983, and constructed prior to July 1, 1985, and a special form encapsulation designed in accordance with the requirements of S 71.4 of this part in effect on June 30, 1989, and constructed prior to July 1, 1991 may continue to be used. Any other special form encapsuiation must meet the requirements of this paragraph.

"Specific activity" of a radionuclide means the radioactivity of the radionuclide per unit mass of that nuclide. The specific activity of a material in which the radionuclide is essentially uniformly dis-tributed is the radioactivity per unit mass of the material.

"State" means the 50 States of the Union, the District of Columbia, the Commonwealth of Puerto Rico, the Virgin Islands, Guam, American Samoa, the Trust Territory of the Pacific Islands, and the Commonwealth of the Northern Mariana Islands.

39

"Surface Contaminated Object (SCO)" means a solid object which is not itself radioactive having radioactive material not exceeding the following limits distributed on its surfaces. SCO must be in one of two groups:

(1) SCO-I Asolidobjectonwhich:

(1) The non-fixed contamination on the accessible surface averaged 2

over 300 cm2 (or the area of the surface if less than 300 cm ) does not exceed 1.08 x 10'4 pCi/cm 2 (4 Bq/cm2 ) for beta and gamma emitters, or 1.08 x 10'5 pCi/cm2 (0.4 Bq/cm 2 ) for alpha emitters; (ii) The fixed contamination on the accessible surface averaged over >

300 cm2 (or the area of the surface if less than 300 cm 2) does not exceed 1.08 pCi/cm 2 (4 x 104 Bq/cm2 ) for beta and gamma emitters, or 0.108 pCi/cm2 (4 x 108 Bq/cm 2 ) for alpha emitters; and L

(iii) The non-fixed contamination plus the fixed contamination on the inaccessible surface averaged over 300 cm2 (or the area of the wrface if less than 300 cm2 ) does not exceed 1.08 pCi/cm2 (4 x 104 Bq/cm 2 ) for beta and gamma emitters, or 0.108 pCi/cm 2 (4 x 108 Bq/cm 2 ) for alpha emitters.

(2) SCO-II t

i A solid object on which the limits for SCO-I are exceeded and on i

which: l 1  ;

(i) The non-fixed contamination on the accessible surface averaged t

over 300 cm2 (or the area of the surface if less than 300 cat) does not exceed 1.08 x 10-2 pCi/cm2 (400 Bq/cm2 ) for beta and gamma emitters or 1.08 x 10'3 pCi/cm 2 (40 Bq/cm 2 ) for alpha emitters;

{

1 40 l

l

. .- - - , . . , , . - . - . - -- -- . - - - - , - - - . . . . .- - -, - ----.-~.-~--- ---.---l

(ii) The fixed contamination on the surface averaged over 300 cm2 (or the area of the surface if less than 300 cm )2 does not exceed 21.6 pCi/cm 2 (8 x 105 Bq/cm2 ) for beta and gamma emitters, or 2.16 pCi/cm2 (8 x 104 Bq/cm 2 ) for alpha emitters; and (iii) The non-fixed contamination plus the fixed contamination on 1

the inaccessible surface averaged over 300 cm2 (or the area of the sur-face if less than 300 cm )2 does not exceed 21.6 pCi/cm2 (8 x 105 Bq/cm2 )

for beta and gamma emitters, or 2.16 pCi/cm2 (8 x 104 Bq/cm 2 ) for alpha '

emitters.

1 "Transport index" means the dimensionless number (rounded up to the

! first decimal place) placed on the label of a package to designate the degree of control to be exercised by the carrier during transportation.

The transport index is determined as follows:  :

(1) For nonfissile material packages, the number expressing the i

, maximum radiation level in millirem per hour at 1 meter from the external surface of the package; or (2) For fissile material packages, the number expressing the maxi-mum radiation level in millirem per hour at 1 meter from the external l surface of the package, or the number determined under S 71.59, whichever number is larger.

"Type A quantity" means a quantity of radioactive material, the aggregate radioactivity of which does not exceed A t for special form  !

radioactive material or As for normal form radioactive material, where At and As are given in Table A-1 of this part or may be determined by procedures described in Appendix A of this part.

"Type B quantity" means a quantity of radioactive material greater i

than a Type A quantity, i

) 41 i

)

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[7590-01]

"Uranium - natural, depleted, enriched" (1) "Natural tranium" means uranium with the naturally occurring distribution of ura.11um isotopes (approximately 0.711 weight percent uranium-235, and trie remainder essentially uranium-238).

(2) "Depleted uranium" means uranium containing less uranium-235 than the naturally occurring distribution of uranium isotopes.

(3) "Enriched uranium" means uranium containing more uranium-235 than the naturally occurring distribution of uranium isotopes.

S 71.5 Transportation of licensed material.

(a) Each licensee who transports licensed material outside of the confinesofitsplantorotherplaceofuse,orwhodeliverslicensed material to a carrier for transport, shall comply with the applicable requirements of the 00T regulations in 49 CFR Parts 170-189 appropriate to the mode of transprt.

(1) Th. licensee shall particularly note DOT regulations in the following areas:

(i) Packaging - 49 CFR Part 173: Subparts A and B and SS 173.401 -

173.478.

(ii) Marking and labeling - 49 CFR Part 172: Subpart D, SS 172.400 -

172.407, and SS 172.436 - 172.440.

(iii) Placarding - 49 CFR Part 172: SS 172.500 - 172.519, 172.556, and Appendices B and C.

(iv) Accident reporting - 49 CFR Part 171: Sb 171.15 and 171.16.

(v) Shipping papers - 49 CFR Part 172: Subpart C.

(2) The licensee shall also note DOT regulations pertaining to the following modes of. transportation:

I 42

[7590-01]

(i) Rail - 49 CFR Part 174: Sut parts A - O and K.

(ii) Air - 49 CFR Part 175 (iii) Vessel - 49 CFR Part 176: Subparts A - F and M.

(iv) Public Highway - 49 CFR Part 177.

(b) If DOT regulations are not applicable to a shipment of licensed material by rail, highway, or water because the shipment or the trans-portation of the shipment is not in interstate or foreign comerce, or t

to a shipment of licensed material by air because the shipment is not transported in civil aircraft, the licensee shall conform to the standards and requirements of the 00T specified in paragraph (a) of this section to the same extent as if the shipment or transportation were in interstate or foreign comerce or in civil aircraft. A request for modification, waiver, or exemption from those requirements, and any notification referred to in those requirements, must be filed with or made to the Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555. ,

I Subpart B - Exemptions ,

L

$ 71.6 [ Reserved.]

i S 71.7 Specific exemptions.

I On application of any interested person or on its own initiative,

the Comission may grant any exemption from the requirements of the l regulations in this part that it determines is authorized by law and i will not endanger life or property or the comon defense and security, j I l

$ 71.8 [ Reserved.]

43 i

.- (7590-01]

6 11.9 Exemption of physicians.

Any physician licensed by a State of the United States to dispense drugs in the practice of medicine is exempt from 5 71.5 with respect to transport by the physician of licensed material for use in the practice of medicine. However, any physician operating under this exemption must be licensed under 10 CFR Part 35 or the equivalent Agreement State regulations.

1 5 71.10 Exemption for low-level materials.

(a) A licensee is exempt from all requirements of this part with re s.. ' to shipment or carriage of a package containing radioactive mate-rid M 3 a specific activity not greater than 0.002 pCi/g (74 KBq/kg).

/4 censee is exempt fror. all requirements of this part, other than S 71.5 and S 71.88, with respect to shipment or carriage of the following packages, provided the packages contain no fissile material or the fissile material exemption standards of 5 71.53 are satisfied:

j (1) A package containing no more than a Type A quantity of radio-active material; (2) A package in which the only radioactive material is low-specific-  ;

activity (LSA) material or surface contaminated objects (SCO), provided the quantity of radioactive material in that package does not exceed 2A . 3  ;

(c) A licensee is exempt from all requirements of this part, other than S 71.5 and S 71.88, with respect to shipment or carriage of low-1 specific activity (LSA) material in group LSA-I and surface contaminated '

objectr (SCO) in group SCO-I. l 5 71.11 [ Reserved.]

44

l- [,7590-01]

Subpart C - General Licenses S 71.12 General license: NRC-approved package. ,

(a) A general license is hereby issued to any licensee of the Commission to transport, or to deliver to a carrier for transport, licensed material in a package for which a license, certificate of com-pliance, or other approval has been issued by the NRC.

(b) This general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the I provisions of Subpart H of this part.

(c) This general license applies only to a licensee who--

(1) Has a copy of the certificate of compliance, or other approval of the package, and has the drawings and other documents referenced in the i

approval relating to the use and maintenance of the packaging and to the actions to be taken prior to shipment; (2) Complies with the terms and conditions of the license, certi-

ficate, er other approvni, as applicable, and the applicable requirements f

q of Subparts A, G, and H of this part; and I

(3) Submits in writing to the Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555, prior to the licensee's first use of the package, the licensee's name and license number and the package identification number specified 1

in the package approval, j (d) This general license applies only when the package approval authorizes use of the package under this general license.

(e) For a Type B or fissile material package, the design of which $

was approved by NRC prior to (the effective date of this regulation), the f general license is subject to the additional restrictions of S 71.13.  !

. I 45 l

[7590-Ok]

S 71.13 Previously approved package.

(a) A Type B package previously approved by the NRC but not desig- i nated as B(U) or B(M) in the identification number of the NRC Certificate of Compliance may be used under the general license of S 71.12 with the following addition 11 conditions:

(1) Fabrication of the packaging was satisfactorily completed by August 31, 1986, as demonstrated by application of its model number in accordance with $ 71.85(c);

(2) A package used for a shipment to a location outside the United States is subject to multilateral approval as defined in S 173.403(o) of  ;

I 00T regulations in 49 CFR Part 173; and (3) A serial number which uniquely identifies each packaging which conforms to the approved design is assigned to and legibly and durably marked on the outside of each packaging.

(b) A Type B(U) package, a Type B(M) package, or a fissile material

package, previously approved by the NRC but without the designation "-85" in the identification number of the NRC Certificate of Compliance, may be f i

used under the general license of 5 71.12 with the following additional l

)

i conditions: l (1) Fabrication of the package is satisfactorily completed by '

December 31, 1995, as demonstrated by application of its model number in accordance with 5 71.85(c);

]

(2) A package used for a shipment to a location outside the United States after December 31, 1992, is subject to multilateral approval as

! defined in S 713.403(o) of DOT regulations in 49 CFR Part 173; and

) (3) After December 31, 1990, a serial number which uniquely iden-tifies each packaging which conforms to the approved design is assigned to and legibly and durably marked on the outside of each packaging.

46

.. .. (7590-01]

(c) The NRC will approve modifications to the design and authorized contents of a Type B package, or a fissile material package, previously approved by the NRC provided--

(1) The modifications are not significant with respect to the design, operating characteristics, or safe performance of the containment system when the package is subjected to the tests specified in SS 71.71 and 71.73; i and (2) The modification to the package satisfies the requirements of -

this part. [

4 (d) The NRC will revise the package identification number to i designate previously approved package designs as B(U) or B(M), as appro-priate, and with the identification number suffix "-85" after receipt of an application demonstrating that the design meets the requirements of l this part.

]

$ 71.14 General license: DOT specification container, r t

(a) A general license is issued to any licensee of the Commission to transport, or to deliver to a carrier for transport, licensed material

! in a specification container for fissile material or for a Type B '

quantity of radioactive material as specified in the regulations of DOT  ;

t in 49 CFR Parts 173 and 178.

(b) This general license applies only to a licensee who has a i

quality assurance program approved by the Commission as satisfying the

]

provisions of Subpart H of this part.

(c) This general license applies only to a licensee who--

i j (1) Has a copy of the specification; and l l

(2) Complies'with the terms and conditions of the specification l l and the applicable requirements of Subparts A, G, and H of this part.

[ 47

[7590-01)

(d) Thisgenerallicenseissubjecttothelimitationthatthe specification container may not be used for a shipment to a location outside the United States after August 31, 1986, except by multilateral approval as defined in S 173.403(o) of DOT regulations in 49 CFR Part 173.

S 71.16 General License: Use of foreign approved package.

(a) A general license is issued to any licensee of the Commission to transport, or to deliver to a carrier for transport, licensed material in a package the JMign of which has been approved in a foreign national competent authority certificate which has bien revalidated by DOT as meet-ing the applicable requirements of 49 CFR 171.12.

(b) Except as otherwise provided in this section, the general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the applicable provisions of Subpart H of this part.

(c) This general license applies only to shipments made to or from locations outside the United States.

(d) This general license applies.only to a licensee who--

(1) Has a copy of the applicable certificate, the revalidation, and the drawings and other documents referenced in the certificate relating to the use and maintenance of the packaging and to the actions to be taken prior to shipment; and (2) Complies with the terms and conditions of the certificate and revalidation and with the applicable requirements of Subparts A, G, and ,

H of this part. With respect to the quality assurance provisions of Sub-part H of this part, the license is exempt from design, construction, and f

fabrication cons,derations.

48 4

-,_ _ .,. ._.-,_.-.___%._- . . - . - __,.._,,,__,-..-.7.

.. (7590-01)

$ 71.18 General license: Fissile material, limited quantity per package.

(a) A general license is issued to any licensee of the Commission l l

to transport fissile material, or to deliver fissile material to a carrier for transport, without complying with the package standards of Subparts E and F of this part if the material is shipped in accordance with this section.

(b) The general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the proiisions

, of Subpart H of this part.

1

(c) This general license applies only when a package contains no more than a Type A quantity of radioactive material, including only one of the following

(1) Up to 40 g of uranium-235; (2) Up to 30 g of uranium-233; (3) Up to 25 g of the fissile radionuclides of plutonium, except that for encapsulated plutonium-beryllium neutron sources in special form, an Ai quantity of plutonium may be present; or (4) A combination of fissile radionuclides in which the sum of the ratios of the amount of each radionuclide to the corresponding maximum amounts in paragraphs (c)(1), (2), and (3) of this section does not exceed unity.

(d) This general license applies only when, except as specified below for encapsulated plutonium-beryllium sources, a package containing more than 15 g of fissile radionuclides is labeled with a transport index not less than the number given by the following equation, where the pack-age contains x g of uranium-235, y g of uranium-233, and z g of the fissile radionuclides of plutonium:

(7590-01]

Minimum Transport Index = (0.40x + 0.67y + z) (1 x ,15. g),

For a package in which the only fissile material is in the form of encap-sulated plutonium-beryllium neutron sources in special form, the transport index based on criticality considerations may be taken as 0.026 times the number of grams of the fissile radionuclides of plutonium in excess of 15 g. In all cases, the transport index must be rounded up to one decimal e place and may not exceed 10.0.

S 71.20 General license: Fissile material, limited moderator per package.

(a) A general license is issued to any licensee of the Commission to transport fissile material, or to deliver fissile material to a carrier for transport, without complying with the package standards of Subparts E and F of this part if the material is shipped in accordance with this section.

(b) The general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the provisions of Subpart H of this part.

(c) This general license applies only when--

(1) The package contains no more than a Type A quantity of radioactive l material; (2) Neither beryllium nor hydrogenous material enriched in deuterium i

is present; (3) The total mass of graphite present does not exceed 150 times j the total mass of uranium-235 plus plutonium; l (4) Substances having a higher hydrogen density than water, e.g.,

certain hydrocarbon oils, are not present, except that polyethelene may be used for packing or wrapping; 50 l

r

. (7590-01]

(5) Uranium-233 is not present, and the amount of plutonium does not exceed 1 percent of the amount of uranium-235; (6) The amount of uranium-235 is limited as follows:

(i) If the fissile radionuclides are not uniformly distributed, the maximum amount of uranium-235 per package may not exceed the value given in Table I of this part; or (ii) If the fissile radionuclides are distributed uniformly (i.e.,

cannot form a lattice arrangement within the packaging), the maximum amount of uranium-235 per package may not exceed the value given in Table II nf this part; and (7) The transport index of each package based on criticality :on-siderations is taken as 10 times the number of grams of uranium-235 in the package divided by the maximum allowable number of grams per package in accordance with Table I or Table II of this part, as applicable.

$ 11.22 General license: Fissile material, limited quantity, controlled shipment.

(a) A general lir.ense is issued to any licensee of the Commission to transport fissile material, or to deliver fissile material to a carrier for transport, without complying with the package standards of Subparts E and F of this part if limited material is shipped in accordance with this section.

(b) The general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the provisions of Subpart H of this part.

(c) This general license applies only when a package contains no more than a Type A quantity of radioactive material and no more than 51

, , (7590-01) 4 Table ! Permissible Mass of Uranium-235 Per Fissile  ;

Material Package Applicable to S 71.20(c)(6)(i) '

(Nonuniform Distribution) 4 Uranium enrichment ,

in weight percent '

of uranium-235 not Permissible maximum grams  ;

exceeding of uranium-235 per package  ;

24 40 i

)'

20 42 l 15 45 11 48 l l 10 51  !

i 9.5 52 l 9 54 8.5 55 i 8 57  ;

, 7.5 59  !

! 7 60 l

, 6.5 62 i 6 65 l 5.5 68 5 72 4.5 76 i 4 80 l 3.5 88 l l

3 100  !

' 2,5 120 i 2 164 i

, 1. 5 272 l 1.35 320 i 1 680 l 0.92 1,200 i Table II Permissible Mass of Uranium-235 Per Fissile '

j Material Package Applicable to S 71.20(c)(6)(ii) ,

I (Uniform Distribution) i Uranium enrichment  !

in weight percent '

i of uranium-235 not Permissible maximum grams exceeding of uranium-235 per package  ;

4 84 j

) 3.5 92 ,

j 3 112  !

2.5 148 i 4

2 240  !

! 1.5 560 i j 1.35 800 l 4

t I

52 l

3

400 g total of the fissile radionuclides of plutonium ericapsulated as plutonium-beryllium neutron sources in special form.

(d) This general license applies only when the fissile radionuclides in the shipment exceed none of the following:

(1) 500 g of uranium-235; (2) 300 g total of uranium-233, and the fissile radionuclides of plutonium; (3) A total quantity of uranium-233, uranium-235, and the fissile radionuclides of plutonium so that the sum of the ratios of the quan-tity of each radionuclide to the quantity specified in paragraphs (d)(1) and (d)(2) of this section exceeds unity; or (4) 2,500 g total of the fissile radionuclides of plutonium encap-sulated as plutonium-beryllium neutron sources in special form.

(e) This general license applies only when shipment of these pack-ages is made under procedures specifically authorized by DOT in accordance with 49 CFR Part 173 of its regulations to prevent loading, transport, or storage of these packages with other fissile eterial shipments.

S 71.24 General license: Fissile material, limited moderator, controlled shipment.

(a) A general license is issued to any licensee of the Commission to transport fissile material, or to deliver fissile material to a carrier for transport, without complying with the package standards of Subparts E and F of this part, if limited material is shipped in accordance with this section.

(b) The general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the provisions of Subpart H of this part.

53

. (7590-01)

(c) This general license applies only when-- ,

(1) No package contains more than a Type A quantity of radioactive <

material; (2) The packaging does not incorporate lead shielding exceeding 5 cm in thickness, tungsten shielding, or uranium shielding; (3) Neither beryllium nor hydrogenous material enriched in deuterium is present; ,

(4) The total mass of graphite present does not exceed 150 times the total mass of uranium-235 and plutonium;

, (5) Substances having a higher hydrogen density than water, e.g.,

certain hydrocarbon oils, are not present, except that polyethylene may be used for packing or wrapping; ,

(6) For fissile contents containing no uranium-233 and less than 1 percent total plutonium if the fissile radionuclides are--

j (i) Not uniformly distributed, the maximum amount of uranium-235 per consignment does not exceed the value given in Table III of this part; or (ii) Distributed uniformly and cannot form a lattice arrangement within the packaging, the maximum amount of uranium-235 per shipment does l not exceed the value given ir Table IV of this part; (7) For fissile contents containing uranium-233 or more than 1 per-

cent plutonium, the total mass of fissile material per shipment is limited so that the sum of the number of grams of uranium-235 divided by 400, the number of grams of plutonium divided by 225, and the number of grams of 4

i 6 4

! 54 1

- - - - - - - - - . - - , - - . _ - - . - - , - - - - - - - - . . . , , .,.n.- -

---

,- m,~-, - + - - - - - - - - - - - - - -

., [7590-01]

Table III Permissible Mass of Uranium-235 per Fissile Material Shipment Applicable to $ 71.24(c)(6)(i)

(Nonuniform Distribution)

Uranium enrichment in weight percent Permissible maximum grams of uranium-235 not of uranium-235 per exceeding consignment 20 520 15 560 11 600 '

10 640 9.5 655 .

9 675 8.5 690 8 710 4

7.5 730 7 750 6.5 780 6 810 5.5 850 t 5 900 4.5 950 4 1,000  ;

3.5 1,100 3 1,250 2.5 1,500  :

2 2,050 ,

1. 5 3,400 1.35 4,000 1 8,500 0.92 15,000 Table IV Permissible Mass of Uranium-235 Per Fissile Material Shipment Applice51e to S 71.24(c)(6)(ii)

(Uniform Distribution)

I Uranium enrichment in weight percent Permissible maximum grams of uranium-235 not of uranium-235 per exceeding consignment 4 1,050 3.5 1,150 l

3 1,400 -

2.5 1,800 2 3,000 1.5 7,000 1.35 10,000 55

. [7590-01) l uranium-233 divided by 250 does not exceed unity as expressed in the  !

formula grams uranium-235 , grams plutonium arams uranium-233 < y.'

400 g 225 g 250 g -

(8) The transport must be direct to the consignee without any intermediate transit storage; and (9) Shipment of these packages is made under procedures specifically authorized by DOT in accordance with 49 CFR Part 173 of its regulations to prevent loading, transport, or storage of these packages with other fissile material shipments.

Subpart 0 - Application for Package Approval 6 71.31 Contents of application.

(a) An application for an approval under this part must include, for each proposed packaging design, the following infonnation.

(1) A package description as required by S 71.33; (2) A package evaluation as required by S 71.35; and (3) A quality assurance program description as required by $ 71.37 t'

or a reference to a previously approved quality assurance program.

(b) Except as provided in S 71.13, an application for modification ,

of a package design, whether for modification of the packaging or author-I ized contents, must include sufficient information to demonstrate that the proposed design satisfies the package standards in effect at the time the application is filed.

(c) The applicant shall identify any established codes and standards proposed for use in package design, fabrication, assembly, testing, 56

.  : , [7590-01) maintenance, and use. In the absence of any codes and standards, the applicant shall describe and justify the basis and rationale used to formulate the package quality assurance program. (

t

$ 71.33 Package description.

The application must include a description of the proposed package ,

l in sufficient detail to identify the package accurately and provide a i t.

sufficient basis for evaluation of the package. The description must  ;

include-- l (a) With respect to the packaging--

(1) Classification as Type B(U), Type 8(M), or fissile material packaging; (2) Gross weight; [

(3) Model number; t

1 i

(4) Identification of the containment system; (5) Specific materials of construction, weights, dimensions, and

] fabrication methods of--

1 l (i) Receptacles; (ii) Materials specifically used as nonfissile neutron absorbers l l or moderators; i

l (iii) Internal and external structures supporting or protecting receptacles; (iv) Valves, sampling ports, lifting devices, and tie-down devices; ,

and u

(v) Structural and mechanical means for the transfer and dissipation of heat; and 4

)

l

. * (7590-01) l ,

(6) Identification and volumes of any receptacles containing coolant.

(b) With respect to the contents of the package--

(1) Identification and maximum '.adioactivity of radioactive

! constituents; j (2) Identification and maximum quantities of fissile constituents; l (3) Chemical and physical form; l

(4) Extent of reflection, the amount and identity of nonfissile l

, materials used as neutron absorbers or moderators, and the atomic ratio of moderator to fissile constituents;

]

4

, (5) Maximum normal operating pressure; i (6) Maximum weight; (7) Maximum amount of decay heat; and I (8) Identification and volumes of any coolants.

L i

r S 71.35 Package evaluation.

l The application must include the following:

l (a) A dem)nstration that the package satisfies the standards specified in Subparts E and F of this part; (b) For a fissile material package, the allowable number of packages .

J l

which may be transported in the same vehicle in accordance with 5 71.59; ,

I  !

j and

(c) For a fissile material shipment, any proposed special controls i e

! and precautions for transport, loading, unloading, and handling and any '

r

! proposed special controls in the event of an accident or delay.

l I

i l

) 58 L

. i

. - . _ - _ = _ . . - _ - - _ _ . _. - .

(7590-01)

I

$ 71.37 Quality assurance.

(a) The applicant shall describe the quality assurance program (see Subpart H of this part) for the design, fabrication, assembly, testing,  ;

maintenance, repair, modification, and use of the proposed package.

(b) The applicant shall identify any specific provisions of the quality assurance program which are applicable to the particular package design under consideration, including a description of the leak testing procedures.

t I

{ S 71.38 Renewal of a certificate of compliance or quality assurance  !

l program approval.

(a) Except as provided in paragraph (b) of this section, each Certifi-

! cate of Compliance or Quality Assurance Program Approval expires at the I i

ll end of the day, in the month and year stated in the approval.

(b) In any case in which a person, not less than 30 days prior to i

j the expiration of an existing Certificate of Compliance or Quality Assur-ance Program Approval issued pursuant to the part, has filed an applica- {

tion in proper form for renewal of either of those approvals, the existing

! Certificate of Compliance or Quality Assurance Program Approval for which '

l the renewal applic3 tion was filed shall not expire until final action on l i

1 i

the application for renewal has been taken by the Commission, i

! (c) In applying for renewal of an existing Certificate of Compliance or Quality Assurance Program Approval, an applicant must submit a consoli- i dated application which incorporates all changes to its program that are incorporated by reference in the existing approval certificate into as few i referenceable documents as reasonably achievable.

i 1

59

(7590-01) 5 71.39 Requirement for additional information.

The Commission may at any time require additional information in i order to enable it to determine whether a license, certificate of com- ,

pliance, or other approval should be granted, renewed, denied, modified,  ;

suspended, or revoked, j l Subpart E - Package Approval Standards I $ 71.41 Demonstration of compliance.

(a) The effects on a package of the tests specified in 5 71.71

' (Normal Conditions of Transport) and the tests specified in i 71.73 h (Hypothetical Accident Conditions) and S 71.61 (deep immersion test) must be evaluated by subjecting a specimen or scale model to a specific ,

test, or by another method of demonstration acceptable to the Commission, as appropriate for the particular feature being considered, i (b) Taking into account the type of vehicle, the method of securing or attaching the package, and the controls to be exercised by the shipper. (

the Commissio ma,' permit the shipment to be evaluated together with the transporting vehicle. . ,

(c) Environmental and test conditions different from those specified in 55 71.71 and 71.73 may be approved by the Commission if the controls proposed to be exercised by the shipper are demonstrated to be adequate to provide equivalent safety of the shipment. l l

$ 71.43 General standards for all packages. l (a) The smallest overall dimension of a package must not be less I than 10 cm (3.94 in).

60

[7590-01)

(b) The outside of a package must incorporate a feature, such as a seal, that is not readily breakable and that, while intact, would be evidence that the package has not been opened by unauthorized persons.

(c) Each package must include a containment system securely closed by a positive fastening device that cannot be opened unintentionally or by a pressure that may arise within the package.

(d) A package must be of materials and construction that assure that there will be no significant chemical, galvanic, or other reaction among the packaging components, among package contents, or between the packaging components and the packcge contents, including possible reaction resulting from inleakage of water to the maximum credible extent. Account must be taken of the behavior of materials under irradiation.

(e) A package valve or other device, the failure of which would allow radioactive contents to escape, must be protected against unautho-rized operation and, except for a pressure relief device, must be provided with an enclosure to retain any leakage.

(f) A package must be designed, constructed, and prepared for ship-ment so that under the tests specified in 5 71.73 (Normal Conditions of Transport) there would be no loss or dispersal of radioactive contents, no significant increase in external surface radiation levels, and no substantial reduction in the effectiveness of the packaging.

(g) A package must be designed, constructed, and prepared for trans-port so that in still air at 38'C (100*F) and in the shade, no accessible surface of a package would hu<e a temperature exceeding 50*C (122*F) in a nonexclusive use shipment or 85'C (185'F) in an exclusive use shipment.

(h) A package must not incorporate a feature intended to allow con-tinuous venting during transport.

61

$ 71.45 Lifting and tie-down standards for all packages.  ;

(a) Any lifting attachment f. hat is a structural part of a package  !

must be designed with a minimum safety factor of three against yielding when used to lift the packt,a in the intended manner, and it must be designed so that failure of any lifting device under excessive load would not impair the ability of the package b meet other reouirements of this j subpart. Any other structural part of the package which could be used to  !

i lif t the package must be capable of being iandered inoperable for lif ting ,

t the package during transport or must be designed with strength equivalent  !

to that required for lifting attachments.

(b) Tie-down devices:

(1) If there is a system of tie-down devices which is a structural j part of the package, the system must be capable of withstanding, without  !

generating stress in any material of the package in excess of its yield strength, a static force applied to the center of gravity of the package having a vertical component of 2 times the weight of tne package with i its contents, a horizontal component along the direction in which the vehicle travels of 10 times the weight of the package with its contents, and a horizontal component in the transverse direction of 5 times the  ;

weight of the package with its contents.  ;

(2) Any other structural part of ti,e package that could be used .

to tie down the package must be capable of being rendered inoperable for tying down the package during transport, or must be designed with  ;

strength equivalent to that required for tie-down devices, j (3) Each tie-down device which is a structural part of a package f

must be designed so that failure of the device under excessive load would ,

l t

i 62 l I

l

- (75i001) *

)  !

not impair the ability of the package to meet other requirements of this i part.

l

$ 71.47 External radiation standards for all packages.

1

A package must be designed and prepared for shipment so that the ,

~

radiation level does not exceed 200 ares /h (2 mSv/h) at  ;

any point on the accessible external surface of the package and the  ;

l transport index defined in $ 71,4 does not exceed 10. For a package i transported as exclusive use by rail, highway, or water, radiation levels 1 i j external to the package may exceed these limits, but must not exceed any l

of the following limits:

f l (a) Radiation levels on the accessible external surface of the I package must not exceed 200 mrea/h (2 mSv/h) unless the following condi- [

l tions are met, in which case the limit is 1000 area /h (10 mSv/h):  :

(1) The shipment is made in a closed transport vehicle; (2) Provisions are made to secure the package so that its position I

within the vehicle remains fixed during transportation; and [

i (3) There are no loading or unloading operations between the beginning  ;

I i d

and end of the transportation; l

! t (b) Radiation levels at any point on the outer surface of the vehicle j i  !

must not exceed 200 arem/h (2 mSv/h), including the upper and lower j 4

l

}a surfaces, or, in the case of a flat-bed style vehicle, at any point on the j

{ vertical planes projected from the outer edges of the vehicle, on the upper l

. l

{ surface of the load (or enclosure if used), and on the lower external surface

{

of the vehicle;  !

I f I

i 63 I

l t

(7530-08]

(c) Radiation levels at any point 2 m from the outer lateral surfaces of the vehicle (excluding the top and underside of the vehicle),

or, in the case of a flat-bed style vehicle, at any point 2 m from the vertical planes projected try tne outer edges of the vehicle (excluding the top and underside of the vehicle) must not exceed 10 mrem /h .

(0.1 mSv/h); and (d) Radiation levels in any normally occupied positions of the vehicle must not exceed 2 mrem /h (.02 mSv/h), except that this provision does not i

apply to private carriers when persons occupying these positions are pro-vided with special health supervision, personnel radiation exposure monitoring devices, and training in accordance with $ 19.12 of this chapter.

l S 71.51 Additional requirements for Type B packages.

(a) Except as provided in 5 71.52, a Type B package, in addition i to satisfying the requirements of $$ 71.41-71.47, must be designed, con-i structed, and prepared for shipment so that under the tests specified in--

(1) Section 71.71 (Normal Conditions of Transport), there would be l

no loss or dispersal of radioactive contents, as demonstrated to a sensi- '

1 l tivity of 10 ' As per hour, no significant increase in external surface  :

1 i radiation levels, and no substantial reduction in the effettiveness of the packaging; and i  !

(2) Section 71.73 (Hypothetical Accident Conditions), there would l be no escape of krypton-85 exceeding 10A in one we6k, no escape of other radioactive material exceeding a total amount As in one week, and no I external radiation dose rate exceeding I rem /h (10 mSv/h) at 1 m from the external surface of the package.

d 64 f

[7590-01) ,

(b) Compliance with the permitted activity release limits of para-graph (a) of this section must not depend upon filters or upon a mechanical cooling system.

$ 71.52 Exemption for low-specific-activity (LSA) packages.

(a) A package need not satisfy tne requirement of 5 71.51(a)(1)  ;

which limits the loss or dispersal of radioactive contents if it containt only low-specific-activity material and is transported as exclusive use, but is subject to $$ 71.43-71.47 of this part, including $ 71.43(f).

(b) A package need not satisfy the requirements of 5 71.51 if it i

contains only low-specific-activity material and is transported as i exclusive use, but is subject to $$71.41 through 71.47 of this part, including $ 71.43(f). This paragraph (b) expires on (one year after  !

effective date).  !

l 5 71.53 Fissile material exemptions, j i

The following packages are exempt from fissile material classifica- r l

tion and from the fissile material standards of 5 71.55 and $ 71.59, but l are subject to all other requirements of this part: I I

(a) A package containing not more than 15 g of fissile material. l If material is transported in bulk, the quantity limitation applies to the conveyance; (b) A package containing homogeneous hydrogenous solutions or l mixtures where:

(1) The minimum ratio of the number of hydrogen atoms to the number  !

I of atoms of fissile radionuclides (H/X) is 5,200; l t

65 I

w-,

[7590-01]  ;

1 (2) The maximum concentration of fissile radionuclides is 5 g/1; and (3) The maximum mass of fissile radionuclides in the package is 800 g, except for a mixture where the total mass of plutonium and uranium-233 exceeds one percent of the mass of uranium-235, the limit is 500 g. If the material is transported in bulk, other than by air-craft, the quantity limitations apply to the conveyanc ;

(c) A package containing uranium enriched in uranium-235 to a maxi-mum of 1 percent by weight, and with a total plutonium and uranium-233 content of up to 1 percent of the inass of urs, fum-235, if the fissile radionuclides are distributed homogeneously throughout the package con-tents and do not form a lattices arrangement within the package; (d) A package containing any fissile material if it does not con-tain more than 5 g of fissile radionuclides in any 10 1 volume, and if the material is packaged so as to maintain this limit of fissile radio-nuclide concentration during normal transport; (e) A package containing not more than 1 kg of plutonium of which not more than 20 percent by mass may consist of plutonium-239, plutor hm-241, or any combination of those radionuclides; or (f) A package containing liquid solutions of uranyl nitrate enriched in uranium-235 to a maximum of 2 percent by weight, with total plutonium and uranium-233 not more than 0.1 percent of the mass of uraniun-235 and with a minimum nitrogren-to-uranium atomic ratio (N/U) of 2.

S 71.55 General requirements for fissile material packages.

(a) A package used for the shipment of fissile material must be designed and constructed in accordance with SS 71.41-71.47. When required 66

[7590-01]

by the total amount of radioactive material, a package used for the ship-ment of ft:sile material must also be designed and constructed in accordance with 8 71.51.

• (b) Except as provided in paragraph (c) of this section, a package used for the shipment of fissile material must be so designed and con-structed and its contents so limited that it would be subcritical if water were to leak into the containment system or liquid contents were to leak out of the containment system so that, under the following conditions, maximum reactivity of the fissile material would be attained:

(1) The must reactive credible configuration consistent with the chemical and physical form of the material; (2) Moderation by water to the most reactive credible extent; and (3) Close full reflection of the containment ;ystem by water on all sides.

(c) The Commission may approve exceptions to the requirements of para. graph (b) of this section if the package incorporates special design features that ensure that no single packaging error would permit leakage, and if appropriate measures are taken before each shipment to ensure that the containment system does not leak.

(d) A package used for the shipment of fissile material must be so designed and constructed and its contents so limited that under the tests specified in S 71.71 (Normal Conditions of Transport)--

(1) The contents would be suberitical; (2) The geometric form of the package contents would not be substantially altered; (3) There would be no leakage of water into the containment system unless, in the evaluation of undamaged packages under S 71.59(b)(1), it 67

.. [7590-01) has been assumed that moderation is present to such an extent as to cause maximum reactivity consistent with the chemical and physical form of the material; and (4) There will be no substantial reduction in the effectiveness of the packaging, including:

(i) No more than 5 percent reduction in the total effective volume of the packaging on which nuclear safety is assessed; (ii) No more than 5 percent reduction in the effective spacing between the fissile contents and the outer surface of the packaging; and (iii) No occurrence of an aperture in the outer surface of the i

packaging large enough to permit the entry of a 10 cm (4 in) cube.

(e) A package used for the shipment of fissile material must be so designed and constructed and its contents so limited that under the tests specified in S 71.73 (Hypothetical Accident Conditions), the package would be suberitical. For this dett .ation, it must be assumed that:

(1) The fissile material is in the most reactive credible configura-tion consistent with the damaged condition of the package and the chemical and physical form of the contents; (2) Water moderation occurs to the most reactive credible extent consistent with the damaged condition of the package and the chemical and physical form of the contents; and (3) There is reflection by water on all sides, as close as is consistent with the damaged condition of the package.

l 1

6 71.59 Standards for arrays of fissile mcterial packages.

(a) A fissile material package must be controlled by either the ,

shipper or the carrier during transport to assure that an array of such 68

• [7590-01]

packages remains subcritical. To enable this control, the designer of a fissile material package shall derive a number "N" based on all the followin.g conditions being satisfied, assuming packages are stacked together in any arrangement and with close reflection on all sides by water:

(1) Five times "N" undamaged packages with nothing between the packages would be subcritical; (2) Two times "N" damaged packages, if each package were subjected to the tests specified in S 71.73 (Hypothetical Accident Conditions) would be subcritical with optimum interspersed hydrogenous moderation; and (3) The value of "N" cannot be less than 0.5.

(b) The transport index based on nuclear criticality control shall be obtained by dividing the number 50 by the value of "N" derived using the procedures specified in paragraph (a) of this section. The value of -

tha transport index for nui: lear cri6icality control may be zero provioed tnat an unlimited number of packages is subcritical such that the value of "N" is effectively equal to infinity under the procedures specified in paragraph (a) of this section. Any transport index greater than zero must be rounded up to the first decimal place.

(c) Where a fissile material package is assigned a nuclear criticality control transport index --

(1) Not in excess of 10, that package may be shipped by any carrier, and that carrier provides adequate criticality control by limiting the sum of the transport indexes to 50 in a non-exclusive use vehicle and to 100 in an exclusi . e vehicle.

69

' [7590-01]

(2) In excess of 10, that package may only be shipped by exclusive use vehicle or other shipper controlled system specified by D0T for fissile material packages. The shipper provides adequate criticality control by limiting the sum of the transport indexes to 100 in an exclusive use vehicle.

S 71.61 Special requirement for irradiated nuclear fuel shipments.

A package for irradiated nuclear fuel with activity greater than

~

106 Ci (37 PBq) must be so designed that if it were immersed under a head of water of at least 200 m, there would be no rupture of the containment system. For demonstration purposes, an external gauge pressure of at least 2 MPa (290 psi) is considered to meet these conditions.

S 71.63 Special requirements for plutonium and other high toxicity radionuclide shipments.

(a) Radioactive materials with an A2 value of 0.01 Ci (0.37 GBq) or less being shipped in a quantity in excess of 20 Ci (0.74 TBq) per package must be shipped as a solid.

(b) Radioactive materials with an.A2 value of 0.01 Ci (0.37 GBq) or less being shipped in a quantity in excess of 20 Ci (0.74 T8q) per pack-age must be packaged in a separate inner container placed within outer packaging that meets the requirements of Subparts E and F for packaging of material in normal form. If the entire package is subjected to the tests specified in S 71.71 (Normal Conditions of Transport), the separate inner container must not release its contents as demonstrated to a sensi-tivity of 10 8 A2/h. If the entire package is subjected to the tests specified in S 71.73 (Hypothetical Accident Conditions), the separate inner container must restrict the loss of its contents to not more than A 2 in 1

70

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one week. Solid radioactive materials in the following forms are exempt from the requirements of this paragraph:

(1) Reactor fuel elements; (2) Metal or metal alloy; and (3) Other radioactive material bearing solids that the Commission determines should be exempt from the requirements of this section.

S 71.64 Special requirements for pit.tonium air shipments.

(a) A package for the shipment of plutonium by air subject to S 71.88(a)(4), in addition to satisfying the requirements of SS 71.41-71.63, as applicable, must be designed, constructed, and prepared for shipment so that under the tests specified in--

(1) Section 71.74 (Plutonium accident conditions)--

(i) the containment vessel would not be ruptured in its post-tested condition and the package must provide a sufficient degree of containment to restrict accumulated loss of plutonium contents to not more than an A2 quantity in a period of 1 week; (ii) the external radiation level would not exceed I rem /h at a distance of 2.95 ft (0.9 m) from the surface of the package in its post-tested condition in air; and (iii) a single package and an array of packages is demonstrated to be subcritical in accordance with this part, except that the damaged con- .

dition of the package must be considered to be that which results from the plutonium accident tests in S 71.74 rather than the hypothetical acci-dent tests in S 71.73; and (2) Paragraph 71.74(c), there would be no detectable leakage of water into the containment vessel of the package. ,

1 71

)

'. - : [7590-01] )

\

l (b) With respect to the package requirements of paragraph (a),

there must be a demonstration or analytical assessment showing that--

(1) The results of the physical testing for package qualification would not be adversely affected to a significant extent by--

(i) The presence, during the tests, of the actual contents that will be transported in the package, and (ii) Ambient water temperatures ranging from 0.6 C (+33 F) to 38 C

(+100 F) for those qualification tests involving water, and ambient atmospheric temperatures ranging from -40*C (-40*F) to +54*C (+130*F) for the other qualification tests.

(2) the ability of the package to meet the acceptance standards prescribed for the accident condition sequential tests would not be adversely affected if one or more tests in the sequence were deleted.

S 71.65 Additional requirements.

The Commission may, by rule, regulation, or order, impose require-ments upon any licensee in addition to those established in this part as it deems necessary or appropriate to protect public health or to minimize danger to life or property.

Subpart F - Package and Special Form Tests 2 S 71.71 Normal conditions of transport.

(a) Evaluation. Evaluation of each package design under normal conditions of transport must include a determination of the effect on that design of the conditions and tests specified in this section.

2The package standards related to the tests in this subpart are contained in Subpart E.

72

[7590-01]

Separate specimens may be used for the free drop test, the compression test, and the penetration test if each specimen is subjected to the water spray test before being subjected to any of the other tests.

(b) Initial conditions. With respect to the initial conditions for the tests in this section, the demonstration of compliance with the requirements of this part must be based on the ambient temperature pre-ceding and following the tests remaining constant at that value between

-29*C (-20*F) and +38*C (100 F) which is most unfavorable for the feature under consideration. The initial internal pressure w' thin the containment system must be considered to be the maximum noreal operating pressure, unless a lower internal pressure consistent with the ambient temperature considered to precede and follow the tests is more unfavorable.

(c) Conditions and tests. (1) Heat. An ambient temperature of 38*C (100 F) in still air, and insolation according to the following table:

Insolation Data Form and Location of Surfate Total Insolation for a 12-hour period (a cal /cm')

flat surfaces transported horizontally:

-base none

-other surfaces 800 Flat surfaces not transported herizontally 200 Curved surfaces 400 (2) Cold. An ambient temperature of -40*C (-40*t) in still air and shade. I l

(3) Reduced external pressure. An external pressure of 25 kPa (3.63 lb/in2) absolute. ,

(4) Increased external pressure. An external pressure of 140 kPa (20.3 lb/in2 ) absoiute.

73

[7590-01)

(5) Vibration. Vibration normally incident to transport.

(6) Water spray. A water spray that simulates exposure to rainfall of approximately 5 cm/h (1.97 in/h) for at least one hour.

(7) Free drop. Between 1.5 and 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the conclusion of the water spray test, a free drop through the distance specified below onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected. For fissile material packages, this free drop must be preceded by a free drop from a height of 0.3 m (0.994 ft) on each corner or, in the case of a cylindrical fissile material package, onto each of the quarters of each rim.

Criteria For Free Drop Test (Weight / Distance)

Package Weight Free Drop Distance kilograms (pounds) meters (feet) less than 5,000 (less than 11,000) 1. 2 (3.94) 5,000 to 10,000 (11,000 to 22,000) 0.9 (2.95) 10,000 to 15,000 (22,000 to 33,100) 0. 6 (1.97) more than 15,000 (more than 33,100) 0.3 (.984)

(8) Corner drop. A free drop onto each corner of the package in succession, or in the case of a cylindrical package onto each quarter of each rim, from a height of 0.3 m (0.984 ft) onto a flat, essentially unyield-ing, horizontal surface. This test applies only to fiberboard or wood rectangular packages not exceeding 50 kg (110 lbs) and fiberboard or wood cylindrical packages not exceeding 100 kg (220 lbs).

(9) Compression. For packages weighing up to 5,000 kg (11,000 lbs),

the package must be subjected, for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, to a compressive load applied uniformly to the top and bottom of the package in the posi-tion in which the package would normally be transported. The compressive load must be the greater of the following:

74

[7590-01]

(i) The equivalent of five times the weight of the package; or (ii) The equivalent of 13 kPa (1.89 lb/in )2 multiplied by the vert-ically projected area of the package.

(10) Penetration. Impact of the hemispherical end of a vertical steel cylinder of 3.2 cm (1.26 in) diameter and 6 kg (13.2 lbs) mass, dropped from a height of 1 m (3.28 ft) onto the exposed surface of the package which is expected to be most vulnerable to puncture. The long axis of the cylinder must be perpendicular to the package surface.

S 71.73 Hypothetical accident conditions.

(a) Test procedures. Evaluation for hypothetical accident condi-tions is to be based on sequential applic.ation of the tests specified in this section, in the order indicated, to determine their cumulative eftect on a package or array of packages. An undamaged specimen may be used for the water immersion tests specified in paragraphs (c)(5) and (6) of this section.

(b) Test condition _s. With respect to the initial conditions for the tests, except for the water immersion tests, to demonstrate cor 'iance with the requirements of this part during testing, the ambient air temper-ature before and after the tests must remain constant at that value between -29*C (-20*F) and +38*C (100 F) which is most unfavorable for the feature under consideration. The initial internal pressure within the containment system must be the maximum normal operating pressure unless a lower internal pressure consistent with the ambient temperature assumed I to precede and follow the tests is more unfavorable. .

1 (c) Tests. Tests for hypothetical accident conditions must be con-ducted as follows:

75

. [7590-01]

(1) Free Drop. A free drop of the specimen through a distance of 9 m (29.5 ft) onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected.

(2) Crush. Subjection of the specimen to a dynamic crush test by positioning the specimen on a flat, essentially unyielding, horizontal surface so as to suffer maximum damage by the drop of a 500 kg (1,100 pounds) mass from 9 m (29.5 ft) onto the specimen. The mass must consist of a solid mild steel plate 1 m (3.28 ft) by 1 m and must fall in a horizontal attitude.

The crush test is required only when the specimen has a mass not greater than 500 kg (1,100 lbs), an overall density not greater than 1,000 kg/m3 (62.4 lbs/ft )3based on external dimensions, and radioactive contents greater than 1,000 A2 not as special form radioactive material.

(3) Puncture. A free drop of the specimen through a distance of 1 m (39.4 in) in a position for which maximum damage is expected, onto the upper end of a solid, vertical, cylindrical, mild steel bar mounted on an essentially unyielding, horizontal surface. The bar must be 15 cm (5.91 in) in diameter, with the top horizontal and its edge rounded to a radius of not more than 6 mm (0.236 in) and of a length as to cause maximum damage to the package, but not less than 20 cm (7.87 in) long. The long axis of the bar must be vertical.

(4) Thermal. Exposure of the specimen fully engulfed, except for a single support system, in a hydrocarbon fuel / air fire of sufficient extent and in sufficiently quiescent ambient conditions to provide an average emissivity coefficient of at least 0.9, with an average flame temperature of at least 800*C (1,472*F) for a period of 30 minutes, or any other thermal test which provides the equivalent total heat input to the package and which provides a time averaged environmental temperature of 76

[7590-01]

800 C. The fuel source shall extend horizontally at least 1 m (3.28 ft),

but shall not extend more than 3 m (9.84 ft), beyond any external surface of the specimen, and the specimen shall be positioned 1 m (3.28 ft) above the surface of the fuel source. For purposes of calculation, the surface absorptivity coefficient must be either that value which the package may be expected to possess if exposed to the fire specified or 0.8, whichever is greater; and the convective coefficient must be that value which may be demonstrated to exist if the package were exposed to the fire specified.

Artificial cooling must not be applied after cessation of external heat input and any combustion of materials of construction must be allowed to proceed until it terminates naturally.

(5) Immersion - all packages. A separate, undamaged specimen must be subjected to water pressure equivalent to immersion under a head of water of at least 15 m (50 ft). For test purposes, an external pressure of water of 150 kPa (21.7 lb/in )2 gauge is considered to meet these conditions.

(6) Immersion - fissile material. For fissile material subject to S 71.55, in those cases where water inleakage has not been assumed for criticality analysis, immersion under a head of water of at least 0.9 m (3 ft) in the attitude for which maximum leakage is expected.

S 71.74 Plutonium accident conditions (a) Test conditions - Sequence of tests. A package must be phys-ically tested to the following conditions in the order indicated to determine their cumulative effect.

77

• [7590-01]

(1) Impact at a velocity of not less than 129 m/sec (423 ft/sec) at l a right angle onto a flat, essentially unyielding, horizontal surface, in the orientation (e.g., side, end, corner) expected to result in maximum damage at the conclusion of the test sequence.

(2) A static compressive load of 31,800 kg (70,100 lbs) applied in the orientation expected to result in maximum damage at the conclusion of the test sequence. The force on the package must be developed between a flat steel surface and a 5 cm (1.97 in) wide, straight, solid, steel bar.

The length of the bar must be at least as long as the diameter of the package and the longitudinal axis of the bar must be parallel to the plane of the flat surface. The load must be applied to the bar in a manner that prevents any members or devices used to support the bar from contacting the package.

(3) Packages weighing less than 227 kg (500 lbs) must be placed upon a flat, essential.. unyielding, horizontal surface and subjected to a weight of 227 kg (500 lbs) falling from a height of 3 m (9.84 ft) and striking in the position expected to result in maximum damage at the con-clusion of the test sequence. The end of the weight contacting the pack-age must be a solid probe made of mild steel. The probe must be the shape of the frustum of a right circular cone, 30.5 cm (12 in) long, 20.3 cm (8 in) in diameter at the base, and 2.54 cm (1 in) in diameter at the end.

The longitudinal axis of the probe must be perpendicular to the hori-  ;

zontal surface. For packages weighing 227 kg (500 lbs) or more, the 1

base of the probe must be placed on a flat, essentially unyielding hori- '

zontal surface and the parkage dropped from a height of 3 m (9.84 ft) onto  !

the probe, striking in the position expected to result in maximum damage 1 at the conclusion of the test sequence.

78 i

[7590-01)

(4) The package must be firmly restrained and supported such that its longitudinal axis is inclined approximately 45 to the horizontal.

The area of the package which made first contact with the impact surface in paragraph (1) of this section must be in the lowermost position. The package must be struck at approximately the center of its vertical projection by the end of a structural steel angle section falling from a height of at least 45.7 m (150 ft). The angle section must be at least 1.83 m (6 ft) in length with equal legs at least 12.7 cm (5 in) long and 1.27 cm (0.5 in) thick. The angle section must be guided in such a way as to fall end-on, without tumbling. The package must be rotated approxi-mately 90* about its longitudinal axis and struck by the steel angle sec-tion falling as before.

(5) The package must be exposed to luminous flames from a pool fire of JP-4 or JP-5 aviation fuel for a period of at least 60 minutes. The luminous flames must extend an average of at least 0.914 m (3 ft) and no more than 3.05 m (10 ft) beyond the package in all horizontal directions.

The position and orientation of the package in relation to the fuel must be that which is expected to result in maximum damage at the conclusion of the test sequence. An alternate method of thermal testing may be sub-stituted for this fire test provided that the alternate test is not of shorter duration and would not result in a lower heating rate to the pack-age. A. the conclusion of the thermal test, the package must be allowed to cool naturally or must be cooled by water sprinkling, whichever is expected to result in maximum damage at the conclusion of the test sequence.

(6) Immersion under at least 0.914 m (3 ft) of water.

79

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[7590-01] 1 l

(b) Individual free-fall impact test. 1 (1) An undamaged package must be physically subjected to an impact at a velocity not less than the calculated terminal free-fall velocity at mean sea level at a right angle onto a flat essentially unyielding horizontal surface, in the orientation (e.g., side, end, corner) expected to result in maximum damage.

(2) This test is not required if the calculated terminal free-fall velocity of the package is less than 129 m/sec (423 ft/sec) or if a velocity not less than either 129 m/sec (423 ft/sec) or the calculated terminal free-fall velocity of the package is used in the sequential test of paragraph (a)(1) of this section.

(c) Individual deep submersion test. An undamaged package must be physically submerged and physically subjected to an external water pres-sure of at least 4.14 MPa (600 lbs/in ),2 -

S 71.75 Qualification of special form radioactive material.

(a) Evaluation of the contents of a single package for qualifica-tion as special form must include a det.ermination of the effect on a specimen of those contents of the tests specified in S 71.77.

(1) Specimens (solid radioactive material or capsules) to be tested must be as normally prepared for loading in a single package, with the radioactive material duplicated as closely as practicable.

(2) A different specimen may be used for each of the tests.

(b) The specimen must not break or shatter when subjected to the impact, percussion, or bending tests.

(c) The specimen must not melt or disperse when subjected to the heat test.

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[7590-01]

(d) After each test, leak-tightness or indispersibility of the speci-men must be determined by a method no less sensitive than the following leaching assessment procedure. For a capsule resistant to corrosion by a, w water and which has an internal void volume greater than 0.1 ml (6.10 x 10-3 in3), an alternative to the leaching assessment is a demonstration of leak-tightness of 10 4 torr-1/s (.03 lb-in/s) (based on air at 25*C (77 F) and one atmosphere differential pressure) for solid radioactive content, or 10 8 torr-1/s (3 x 10 4 lb-in/s) for liquid or gaseous radioactive content.

(1) The specimen must be imm6rsed for 7 days in water at ambient temperature. The water must have a pH of 6-8 and a maximum conductivity of 10 pmho/cm at 20*C (68*F). Encapsulated material is not subject to the 7-day requirement.

(2) The water with specimen must then be heated to a temperature of 50*C i 5*C (122* i 9 F) and maintained at this temperature for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

(3) The activity of the water must be determined at that time.

(4) The specimen must then be stored for at least 7 days in still air of humidity not loss than 90 percent and a temperature not less than 30 C (86*F).

(5) The specimen must then be immersed in water having a p" of 6-8 and a maximum conductivity of 10 pmho/cm at 20*C, and the water with spect-men heated to 50 1 5*C (122* i 9*F) and maintained at this temperatcre for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

(6) The activity of the water must be determined at that time.

(7) The activities detennined in paragraphs (d)(3) and (d)(6) of this section must not exceed 0.0541 pCi (2 kBq).

81

[7590-01]

(e) A specimen that comprises or simulates radioactive material contained in a sealed capsule need not be subjected to the leak-tightness procedure specified in this section provided it is alternatively subjected to any of the tests prescribed in the International Organization for Standardization document ISO /TR4826-1979(E), "Sealed radioactive sources -

Leak test methods",4 which are acceptable to NRC.

S 71.77 Tests for special form radioactive material.

(a) Impact test. The specimen must fall onto a flat, horizontal, t

essentially unyielding surface from a height of not less than 9 m (29.5 ft).

(b) Percussion test. The specimen must be placed on a sheet of lead which is supported by a smooth solid surface and struck by ;e flat face of a steel billet so as to produce an impact equivalent to that resulting from a free fall of 1.4 kg (3.09 lbs) through 1 m (39.4 in).

The flat face of the billet must be 25 mm (0.984 in) in diameter with the edges rounded to a radius of 3 1 0.3 mm (0.118 1 0.0118 in). The lead, of hardness number 3.5 to 4.5 on the Vickers scale and not more than 25 mm (0.984 in) thick, must cover an area greater than that covered by tha specimen. A fresh surface of lead must be used for each impact. The billet must strike the specimen so as to cause maximum damage.

(c) Bending test. The test is applicable only to long, slender sources with both a minimum length of 10 cm (3.94 in) and a length to mini-mum width ratio not less than 10. The specimen must be rigidly clamped 1 in a horizontal position so that one half of its length protrudes from the face of the clamp. The orientation of the specimen must ensure that

'Available from Tierican National Standards Institute, 1430 Broadway, New York, NY 10018.

82

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[7590-01]

the specimen will suffer maximum damage when its free end is struck by the flat face of a steel billet. The billet must strike the specimen so as to produce an impact equivalent to that resulting from a free vertical fall of 1.4 kg (3.09 lb) through 1 m (39.4 in). The flat face of the billet must be 25 mm (0.984 in) in diameter with the edges rounded off to a radius of 3 1 0.3 mm (0.118 1 0.0118 in).

(d) Heat test. The specimen must be heated to a temperature of not less than 800*C (1472*F) in an atmosphere which is essentially air, and held at that temperature for a period of 10 minutes and must then be allowed to cool.

(e) A specimen that comprises or simulates radioactive material contained in a sealed capsule need not be subjected to the following:

(1) The impact test and the percussion test of this section provided it is alternatively subjected to the Class 4 impact test prescribed in the International Organization for Standardization document ISO 2919-1980(E),

"Sealed radioactive sources - Classification";5 and (2) The heat test of this see. tion provided it is alternatively sub-jected to the Class 6 temperature test specified in the International Organization for Standardization document ISO 2919-1980(E),"Sealed radioactive sources - Classification."s Subpart G - Operating Controls and Procedures 6 71.81 Applicability of operating controls and procedures.

A licensee subject to this part, who under a general or specific license transports licensed material or delivers licensed material to a 6 Ibid.

61 bid.

83

[7590-01) carrier for transport, shall comply with the requirements of this Sub-part G, with the quality assurance requirements of Subpart H of this part, and with the general provisions of Subpart A of this part.

S 71.83 Assumptions as to unknown properties.

When the isotopic abundance, mass, concentration, degree of irradia-tion, degree of moderation, or other pertinent property of fissile material in any package is not known, the licensee shall package the fissile mate-rial as if the known properties have credible values that will cause the maximum neutron multiplication.

S 71.85 Preliminary determinations.

Prior to the first use of any packaging for the shipment of licensed material--

(a) The licensee shall ascertain that there are no cracks, pinholes, uncontrolled voids, or other defects which could significantly reduce the effectiveness of the packaging; (b) Where the maximum normal operating pressure will exceed 35 kPa (5.08 lb/in )2 gauge, the licensee shall test the containment system at an internal pressure at least 53 percent higher than the maximum normal operating pressure to verify the capability of that system to maintain its structural integrity at that pressure; and (c) The licensee shall conspicuously and durably mark the packaging with its model number, serial number, gross weight, and a package iden-tification number assigned by the Nuclear Regulatory Commission. Prior to applying the model number., the licensee shall determine that the pack-aging has been fabricated in accordance with the design approved by the Commission.

84

. [7590-01]

S 71.87 Routine determinations.

Prior to each shipment of licensed material, the licensee shall ensure that the package with its contents satisfies the applicable requirements of this part and of the license. The licensee shall determine that--

(a) The package is proper for the contents to be shipped; (b) The package is in unimpaired physical condition except for superficial defects such as marks or dents; (c) Each closure device of the packaging, including any required gasket, is properly installed and secured and free of defects; (d) Any system for containing liquid is adequately sealed and has adequate space or other specified provision for expansion of the liquid; (e) Any pressure relief device is operable and set in accordance with written procedures; (f) The package has been loaded and closed in accordance with written procedures; (g) For fissile material, any moderator or neutron absorber, if required, is present and in proper condition; (h) Any structural part of the package which could be used to lift or tie down the package during transport is rendered inoperable for that purpose unless it satisfies the design requirements of 6 71.45; (i)(1) The level of non-fixed (removable) radioactive contamina-tion on the external surfaces of each package offered for shipment is as low as reasonably achievable. The level of non-fixed radioactive con-tamination may be determined by wiping an area of 300 square centimeters of the surface concerned with an absorbent material, using moderate pressure, and measuring the activity on the wiping material. Sufficient measurements must be taken in the most appropriate locations to yield a 85

[7590-01]

representative assessment of the non-fixed contamination levels. Except as provided under paragraph (i)(2) of this section, the amount of radio-activity meast,'ed on any single wiping material when averaged over the surface wiped, must not exceed the limits given in Table V of this part at any time during transport. Other methods of assessment of equal or greater efficiency may be used. When other methods are used, the detec-tion efficiency of the method used must be taken into account, and in no case may the non-fixed contamination on the external surfaces of the i package exceed 10 times the limits listeri in Table V.

Table V Removable External Radioactive Contamination Wipe Limits Maximum Permissible Limits Contaminant Bq/cm3 pC1/cm3 dpm/cm' Beta gamma emitting radionuclides; all radionuclides with half-lives less than 10 days; natural uranium; natural thorium; uranium-235; uranium-238; thorium-232; thorium-228 and thorium-230 when contained in ores or physical concentrates...... 0.40 1.08 x 10.s 22.0 All other alpha emitting radionuclides.. ................... 0.04 1.08 x 10 8 2.2 (2) In the case of packages transported as exclusive use shipments by rail or highway only, the non-fixed radioactive contamination at any time during transport must not exceed 10 times the levels prescribed in paragraph (i)(1) of this section. The levels at the beginning of transport must not exceed the levels prescribed in paragraph (i)(1) of this section; (j) External radiation levels around the package and around the vehicle, if applicable, will not exceed the limits specified in S 71.47 at any time during transportation; and 86

- [7590-01]

(k) Accessible package surface temperatures will not exceed the limits specified in 9 71.43(g) at any time during transportation.

6 71.88 Air transport of plutonium.

(a) Notwithstanding the provisions of any general licenses and notwithstanding any exemptions stated directly in this part or included indirectly by citation of 49 CFR Chapter 1, as may be applicable, the licensee shall assure that plutonium in any form, whether for import, I export, or domestic shipment, is not transported by air or delivered to a carrier for air transport unless:

(1) The plutonium is contained in a medical device designed for individual human application; or (2) The plutonium is contained in a material in which the specific activity is not greater than 0.002 pCi/g (74 Bq/g) of material and in which the radioactivity is essentially uniformly distributed; or (3) The plutonium is shipped in a single package containing no more than an A2 quantity of plutonium in any isotope or form and is shipped in accordance with 6 71.5 of this part; or (4) The plutonium is shipped in a package specifically authorized for the shipment of plutonium by air in the Certificate of Compliance for that package issued by the Commission.

(b) Nothing in paragraph (a) of this section is to be interpreted

. as removing or diminishing the requirements of 9 73.24 of this chapter.

(c) For a shipment of plutonium by air which is subject to paragraph (a)(4), the licensee shall, through specici arrangement with the carrier, require the following operational controls.

87

[7590-01]

1 I (1) A plutonium package weighing less than 40 kg (88 lbs), and having its height and diameter both less than 50 cm (19.7 in), must be stowed aboard the aircraft on the main deck or the lower cargo compart-ment in the af t-most location that is possible for cargo of its size and weight. Any other plutonium package must be stowed aboard the aircraft on the main deck in the aft-most location that is possible for cargo of '

its size and weight. No other type of cargo may be stowed aft of a plutonium package.

(2) A plutonium package must be secured and restrained to prevent shifting under normal transport. A plutonium packaging weighing 40 kg (88 lbs) or more must be securely cradled and tied down to the main deck of the aircraft such that the tiedown system is capable of providing package ,

restraint against the following inertial forces acting separately relative to the deck of the aircraft: Upward, 2g; Forward, 9g; Sideward, 1.5g; Downward, 4.5g.

(3) A plutonium package weighing less than 40 kg (88 lbs), and i having its height and diameter both less than 50 cm (19.7 in), must not be transported aboard a aircraft carrying other cargo bearing the "Explosive A" label. Any other plutonium package must not be transported aboard an aircraft carrying other cargo bearing any of the following  !

hazardous material labels:

Explosive A Explosive B Explosive C Spontaneously Combustible i

3 Dangerous When Wet j l

88 ,

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[7590-01]

2 Organic Peroxide ,

Non-Flamable Gas Flamable Liquid Flamable Solid Flamable Gas 1 i

Oxidizer Corrosive The above restriction does not apply to hazardous material cargo labeled

[

solely as:

Radioactive I Radioactive II Radioactive III Magnetized Materials '

Poison Poison Gas Irritant 1 Etiologic Agent l l l

S 71.89 Opening instructions.

Prior to delivery of a package to a carrier for transport, the I licensee shall ensure that any special instructions needed to safely open the package have been sent to or otherwise made available to the consignee for the consignee's use in acsordance with 6 20.205 of this chapter.

A l 89 i

I j

[7590-01]

, 9 71.91 Records.

(a) Each licensee shall maintain for a period of 3 years after shipment a record of each shipment of licensed material not exempt under S 71.10, showing where applicable-- ,

(1) Identification of the packaging by model number and serial number; (2) Verification that there are no significant defects in the packaging, as shipped; (3) Volume and identification of coolant;

(4) Type and quantity of licensed material in each package, and i the total quantity of each shipment; (5) For each item of irradiated fissile material; (i) Identification by model number and serial number; ,

(ii) Irradiation and decay history to the extent appropriate to .

demonstrate that its nuclear and thermal characteristics comply with l

license conditions; and (iii) Any abnormal or unusual condition relevant to radiation safety, (6) Date of the shipment; f ,

(7) For fissile packages and for Type B packages, any special i

controls exercised;

(8) Name and address of the transferee; (9) Address to which the shipment was made; and I i

(10) Results of the determinations required by 5 71.87.  ;

i; ,

(b) The licensee shall make available to the Commission for '

inspection, upon reasonable notice, all records required by this part.

(c) The licensee shall maintain, during the life of the packaging l t

l l to which they pertain, sufficient quality assurance records to furnish j 90 l

/ [7590-01]

documentary evidence of the quality of packaging components which have safety significance and of services affecting quality. The records to be maintained include results of the determinations required by S 71.85, of monitoring, inspection, and auditing of work performance during the design, fabrication, assembly, testing, modification, maintenance, and repair of the packaging. Whera some person other than the licensee pro-vides these services, a cartif* cation from that person that the services have been ut :i: '.+, that adequate records of them are being maintained will su4ica.

4 it t ;  :< s  :,, . +v t g' Tr w iteer.s. s.'r 'trmis the Commission, at all reasonable l times, to t w !, ,. " . . t:aA mattrial, packaging, premises, and

facilities dr. wh v '..e . isnsed material or packaging is used, produced, tested, stored, or sn gged.

(b) The licem a shall perform, and permit the Commission to perform, any tests tbt Commission deuas necessary or appropriate for the administra-tion of the regulations ir, this chapter.

(c) ine 'ilcensee shall notify the Administrator of the appropriate Nuclear Regulatory Commission Regional Office listed in Appendix A of Part 73 of tnis chapter at least 45 days prior to fabrication of a package to be used for the shipment of licensed material having a decay heat load  ;

in excess of 5 kW or with a maximum normal operating pressure in excess of 103 kPa (14.9 lb/in 2) gauge.

l 4 91

- [7590-01]

6 71.95 Reports.

The licensee shall report to the Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555, within 30 days--

(a) Any instance in which there is significant reduction in the effectiveness of any Type B or fissile approved packaging during use; (b) Details of any defects with safety significance in Type B or fissile packaging after first use, with the means employed to repair the defects and prevent their recurrence; and

. (c) Instances in which the conditions of approval in the certifi-cate of compliance were not observed in making a shipment.

l 5 71.97 Advance notification of shipment of nuclear w ste.  ;

(a) Except as specified in paragraph (b) of this section, prior to the transport or delivery to a carrier for transport of licensed material outside the confines of the licensee's plant or other place of use or storage, each licensee shall provide advance notification to the governor of a State, or the governor's designee, of the shipment to, through, or .

across the boundary of the State.

(b) Advance notification is required only when--

(1) The licensed material is required by this part to be in Type B i

packaging for transportation;  ;

(2) The licensed material other than irradiated fuel is being trans-ported to, through, or across State boundaries to a disposal site or to a collection point for transport to a disposal site; i

(3) The quantity of licensed material in a single package exceeds the smallest of the following:

)

92 l

l

[7590-01]

I (i) 3,000 times the At value of the radionuclides as specified in Appendix A, Table A-1 for special form radioactive material; (ii) 3,000 times the A2 value for the radionuclides as specified in Appendix A, Table A-1 for normal form radioactive material; or (iii) 30,000 Ci(1.11 P8q); and (4) The quantity of irradiated fuel is less than that subject to advance notification requirements of 5 73.37(f) of this chapter.

I' (c) Procedures for submitting advance notification.

(1) The notification must be made in writing to the office of each ,

I appropriate governor or governor's designee and to the Administrator of the appropriate Nuclear Regulatory Commission Regional Office listed in j AppeMix A of Part 73 of this chapter.

! (2) A notification delivered by mail must be postmarked at least 7 days before the beginning of the 7-day period during which departure of the shipment is estimated to occur.

(3) A notification delivered by messenger must reach the office of the governor or of the governor's designee at least 4 days before the beginning of the 7-day period during which departure of the shipment is ,

estimated to occur. ,

(i) A list of the names and mailing addresses of the governors' designees receiving advance notification of transportation of nuclear waste was published in the Federal Register on June 30, 1987

! (52 FR 24357),

i

. (ii) The list will be published annually in the Federal Register

. on er about June 30 to reflect any changes in information.

! (iii) A list of the names and mailing addresses of the governors' i

designees is available upon request from the Director, State, Local and l 93 i

)

(7590-01)

Indian Tribe Programs, U.S. Nuclear Regulatory Comission, Washington, DC 20555.

(4) The licersee shall retain a copy of the notification as a record for 1 year.

(d) Information to be furnished in advance notification of shipment.

Each advance notification of shipment of nuclear waste must contain the following information:

(1) The name, address, and telephone number of the shipper, carrier, and receiver of the nuclear waste shipment; (2) A description of the nuclear waste contained in the shiptret, as specified in the regulations of DOT in 49 Cin Part 172, SS 172.202 and 172.203(d);

(3) The point of origin of the shipment and the 7-day period during which departure of the shipment is estimated to occur; (4) The 7-day period during which arrival of the shipment at state boundaries is estimated to occur; (5) The destination of the shipment, and the 7-day period durir.g which arrival of the shipment is estimated to occur; and (6) A point of contact with a telephone number for current shipment information.

(e) Revision notice. A licensee who finds that schedule information previously furnished to a governor or governor's designee in accordance with this section will not be met shall telephone a responsible individual in the office of the governor of the State or of tha governor's designee and inform that individual of the 1xtent of the delay beyond the schedule originally reported. The licensee shall maintain a record of the name of the individual contacted for 1 year.

94

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[7590-01)

(f) Cancellation notice.

(1) Each licensee who cancels a e clear waste shipment for which advance notification has been sent shall send a cancellation notice to the governor of each State or the governor's designee previously notified and to the Administrator of the appropriate Nuclear Regulatory Commission i Regional Office listed in Appendix A of Part 73 of this chapter, a (2) The licensee shall state in the notice that it is a cancellation and shall identify the advance notification which is being cancelled.

The licensee shall retain a copy of the notice as a record for 3 years.

S 71.99 Violations.

An injunction or other court order may be obtained prohibiting any l violation of any provision of the Atomic Energy Act of 1954, as amended, (the Act) or Title II of the Energy Reorganization Act of 1974, as amended,  ;

i j or any regulation or order issued under the acts. A court order may be i obtained for the payment of a civil penalty imposed under section 234 of the Act for violation of sections 53, 57, 62, 63 81, 82, 101, 103, 104, t

107, or 109 of the Act, or section 206.of the Energy Reorganization Act of 1974, as amended; or any rule, regulation, or order issued under the -

Acts, or any term, condition, or limitation of any license issued under the Acts, or for any violation for which a license may be revoked under [

^

section 186 of the Act. Any person who willfully violates any provision 3

of the Act or any regulation or order issued under the Acts may be guilty of a crime and, upon conviction, may be punished by fine or imprisonment ,

or both, as provided by law.  ;

l 95 t

1 i

Subpart H - Quality Assurance

$ 71.101 Quality assurance requirements.

(a) Purpose. This subpart describes quality assurance requirements l applying to design, purchase, fabrication, handling, shipping, storing, cleaning, assembly, inspection, testing, operation, maintenance, repair, and modification of components of packaging which are important to safety.

As used in this subpart, "quality assurance" comprises all those planned  :

and systematic actions necessary to provide adequate confidence that a I

system or component will perform satisfactorily in service. Quality assur-4 ance includes quality control, which comprises those quality assurance ,

actions related to control of the physical characteristics and quality of i the matorial or component to predetermined requirements.

(b) Estab',ishment of program. Each licensee shall establish, I maintain, and execute a quality ascurance program satisfying each of the I

applicable criteria of this subpart, and satisfying any specific provi-sions which are applicable to the licensee's activities including 1

l procurement of packaging. The licensee shall apply the applicable l criteria in a graded approach and to an extent that is consistent with i 1 ,

I their importance to safety. t

(c) Approval of program. Prior to the use of any package for the  !

shipmentoflicensedmaterialsubjecttothissubpart,eachlicenseeshall i obtain Commission approval of its quality assurance program. Each licensee !

shall file a description of its quality assurance program, including a l j discussion of which requirements of this subpart are applicable and how they j i

t will be satisfied, with the Director, Office of Nuclear Material Safety  ;

I

! and Safeguards U.S. Nuclear Regulatory Commission, Washington, DC 20555. i j l 4

! 96 l

}

. .- [7590-01]

(d) Existing package designs. The provisions of this paragraph deal with packages which have been approved for use in accordance with this part prior to January 1,1979, and which nave been designed in accordance with the provisions of this part in effect at the time of application for package approval. Those packages will be accepted as having been designed in accordance with a quality assurance program which satisfies the provisions of paragraph (b) of this section. ,

(e) Existing packages. The provisions of this paragraph deal with packages which have been approved for use in accordance with this part i prior to January 1, 1979, have been at least partially fabricated prior to that date, and for which the fabrication is in accordance with the provisions of this part in effect at the time of application for approval of package design. These packages will be accepted as having been fabricated and assembled in accordance with a quality assurance program which satisfies the provisions of paragraph (b) of this section.

(f) Previously approved programs. A Commission-approved quality assurance program which satisfies the applicable criteria of Appendix B of Part 50 of this chapter and which is established, maintained, and executed with regard to transport packages will be accepted as satisfying the requirements of paragraph (b) of this section. Prior to first use, i the licensee shall notify the Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555, of its intent to apply its previously approved Appendix B program to trans- ,

portation activities. The licensee shall identify the program by date of i submittal to the Commission, Docket Number, and date of Commission approval.  ;

l l

97

.  : tCWERPupJ

$ 71.103 Quality assurance oracnization.

(a) The licensee 7 shall be responsible for the establishment and execution of the quality assurance program. The licensee may delegate to others, such as contractors, agents, or consultants, the work of estab-lishing and executing the quality assurance program, or any part of the quality assurance program, but shall retain responsibility for the program. The licensee shall clearly establish and delineate in writing the authority and duties of persons and organizations performing activities affecting the safety-related functions of structures, systems, e and components. These activities include performing the functions asso-ciated with attaining quality objectives and the quality assurance functions.

(b) The quality assurance functions are--

(1) Assuring that an appropriate quality assurance program is established and effectively executed; and (2) Verifying, by procedures such as checking, auditing, and inspection, that activities affecting the safety-related functions have been performed correctly.

(c) The persons and organizations performing quality assurance functions must have sufficient authority and organizational freedom to--

(1) Identify quality problesis; (2) Initiate, incommend, or provide solutions; and (3) Verify implementation of solutions.

1 VWhile the term "licensee" is used in these criteria, the requirements are applicable to whatever design, fabrication, assembly, arid testing of the package is accomplished with respect to a package prior to the time a package approval is issued.

98

(7590-01)

(d) The persons and organizations performing quality assurance functions shall report to a management lovel which assures that the required authority and organizational freedom, including sufficient independence from cost and schedule when opposed to safety considerations, are provided.

(e) Because of the many variables involved, such as the number of personnel, the type of activity being performed, and the location or locations where activities are performed, the organizational structure for executing the quality assurance program may take various forms pro-i vided that the persons and organizations assigned the quality assurance functions have the regr. ired authority and organizational freedom.

(f) Irrespective of the organizational structure, the individual (s) assigned the responsibility for assuring effective execution of any portion of the quality assurance program at any location where activities subject to this section are being performed must have direct access to the levels of management necessary to perform this function.

(

l 6 71.105 Quality assurance program.

(a) The licensee shall establish, at the earliest practicable time, consistent with the schedule for accomplishing the activities, a quality assurance program which complies with the requirements of this section.

The licensee shall document the quality assurance program by written pro-cedures or instructions and shall carry out the program in accordance with those procedures throughout the period during which the packaging is used. The licensee shall identify the material and components to be coveredbythequalityassuranceprogram,themajororganizations participating in the program, and the designated functions of these organizations. ,

99

L/bsu-01J (b) The licensee, through its quality assurance program, shall pro-vide control over activities affecting the quality of the identified materials and components to an extent consistent with their importance to safety, and as necessary to assure conformance to the approved design of each individual package used for the shipment of radioactive material.

The licensee shall assure that activities affecting quality are accom-plished under suitably controlled conditions. Controlled conditions include the use of appropriate equipment; suitable environmental condi-tions for accomplishing the activity, such as adequate cleanliness; and assurance that all prerequisites for the given activity have been satisfied. The licensee shall take into account the need for special controls, processes, test equipment, tools, and skills to attain the required quality, and the need for verification of quality by inspection and test.

(c) The licensee shall base the requirements and procedures of its quality assurance program on the following considerations concerning the complexity and proposed use of the packege and its components:

(1) The impact of malfunction or failure of the item to safety; (2) The design and fabrication complexity or uniqueness of the item; (3) The need for special controls and surveillance over processes and equipment; (4) The degree to which functional compliance can be demonstrated by inspection or test; and (5) The quality history and degree of standardization of the item.

(d) The licensee shall provide for indoctrination and training of personnel performing activities affecting quality as necessary to assure 100

, , [7590-01]

that suitable proficiency is achieved and maintained. The licensee shall review the status and adequacy of the quality assurance program at estab-lished intervals. Management of other organizations participating in the quality assurance program shall review regularly the status and adequacy of that part of the quality assurance program which they are executing.

S 71.107 Package design control.

(a) The licensee shall establish measures to assure that applicable regulatory requirements and the package design, as specified in the

, license for those materials and components to which this section applies, are correctly translated into specifications, drawings, procedures, and

instruction

.. These measures must include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from standards are controlled. Measures must be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are

] essential to the safety-related functions of the materials, parts, and components of the packaging.

(b) The licensee shall establish measures for the identification and control of design interfaces and for coordination among participating j

design organizations. These measures must include the establishment of written procedures among participating design organizations for the review, approval, release, distribution, and revision of documents involving desigr. interfaces. The design control measures must provide for verifying or checking the adequacy of design, by methods such as design reviews, alternate or simplified calculational methods, or by a  :

1 i suitable testing program. For the verifying or checking process, the 101

[7590-01) licensee shall designate individuals or groups other than those who were responsible for the original design, but who may be from the same  !

organization. Where a test program is used to verify the adequacy of a specific design feature in lieu of other verifying or checking processes, the licensee shall include suitable qualification testing of a prototype or sample unit under the most adverse design conditions. The licensee shall apply design control measures to items such as the following:

(1) Criticality physics, radiation shielding, stress, thermal, hydraulic, and accident analyses;

, (2) Compatibility of materials; ,

(3) Accessibility for inservice inspection, maintenance, and repair; (4) Features to facilitate decontamination; and (5) Delineation of acceptance criteria for inspections and tests.

(c) The licensee shall subject design changes, including field changes, to design control measures commensurate with those applied to the original design. Changes in the conditions specified in the package approval rquire NRC approval.  ;

S 71.109 Procurement document control.

The licensee shall establish measures to assure adequate quality is

, required in the documents for procurement of material, equipment, and services, whether purchased by the licensee or by its contractors or subcontractors. To the extent necessary, the licensee shall reqvire j contractors or subcontractors to provide a quality assurance program consistent with the applicable provisions of this part.

h 1 l 102 -

-. - - - - , . - , - - , - , , - . , , . - - _,-- - . - - . - - , - , _ , -_ ,- ---___ , -._ .,_ , - - , -,.-. ,-_._ s

[7590-01)

S 71.111 Instructions, procedures, and drawings.

The licensee shall prescribe activities affecting quality by docu-mented instructions, procedures, or drawings of a type appropriate to the circumstances and shall require that these instructions, procedures, and drawings be followed. The instructions, procedures, and drawings must include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

4 5 71.113 Document control.

The licensee shall establish measures to control tne issuance of 4

documents such as instructions, procedures, and drawings, including changes, which prescribe all activities affecting quality. These measures must assure that documents, including changes, are reviewed for adequacy, approved for release by authorized personnel, and distributed i

and used at the location where the prescribed activity is performed.

4 These measures must assure that changes to documents are reviewed and  ;

+

approved.  !

S 71.115 Control of purchased material, equipment, and services.

(a) The licensee shall establish measures to assure that purchased material, equipment, and services, whether purchased directly or through contractors and subcontractors, conform to the procurement documents.

These measures must include provisions, as appropriate, for source evaluation and selection, objective evidence of quality furnished by the  ;

contractor or subcontractor, inspection at the contractor or subcontractor source, and examination of products upon delivery.

1 103 J

~~~

[7590-01)

(b) The licensee shall have available documentary evidence that material and equipment conform to the procurement specifications prior to installation or use of the material and equipment. The licensee shall retain or have available this documentary evidence for the life of the package to which it applies. The licensee shall assure that the evidence is sufficient to identify the specific requirements met by the purchased material and equipment.

I (c) The licensee shall assess the effectiveness of the control of quality by contractors and subcontractors at intervals consistent with the importance, complexity, and quantity of the product or services.

S 71.117 Identification and control of materials, parts, and components.

The licensee shall establish me ures for the identificttion and control of materials, parts, and components. These measures aiust assure that identification of the item is maintained by heat number, part number, or other appropriate means, either on the item or on records traceable to the item, as required throughout fabrication, installation, ,

and use of the item. These identification and control measures must be designed to prevent the use of incorrect or defective materials, parts, and components.

S 71.119 Control of special processes.

The licensee shall establish measures to assure that special pro-cesses, including welding, heat treating, and nondestructive testing, are controlled and accomplished by qualified personnel using qualified pro-cedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.

104

tm

• e'M9J S 71.121 Internal inspection.

The licensee shall establish and execute a program for inspection of activities affecting quality by or for the organization performing the activity to verify conformance with the documented instructions, proce-dures, and drawings for accomplishing the activity. The inspection must be performed by individuals other than those who performed the activity being inspected. Examination, measurements, or tests of material or pro-ducts processed must be performed for each work operation where necessary to assure quality. If direct inspection of processed material or products is not carried out, indirect control by monitoring processing methods, equipment, and personnel must be provided. Both inspection and process monitoring must be provided when quality control is inadequate without both. If mandatory inspection hold points, which require witness-ing or inspecting by the licensee's designated representative and beyond which work should not proceed without the consent of its designated repre-sentative, are required, the specific hold points must be indicated in appropriate documents.

S 71.123 Test control.

The licensee shall establish a test program to assure that all test-ing required to demonstrate that the packaging components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements of this part and the requirements and acceptance limits contained in the package approval. The test procedures must include provisions for assuring that all prerequisites for the given test are met, that adequate test instrutantation is available and used, and that the test is performed i

105

under suitable environmental conditions. The licensee shall document and l evaluate the test results to assure that test requirements have been satisfied.

1 S 71.125 Control of measuring and test equipment.

The licensee shall establish measures to assure that tools, gauges, instruments, and other metsuring and testing devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specified times to maintain accuracy within necessary limits.

S 71.127 Handling, storage, and shipping control.

The licensee shall establish measures to control, in accordance with instructions, the handling, storage, shipping, cleaning, and preservation of materials and equipmt.it to be used in packaging to prevent damage or deterioration. When necessary for particular products, special protective environments, such as inert gas atmosphere, and specific moisture content and temperature levels must be specified and provided.

S 71.129 Inspection, test, and operating status.

(a) The licensee shall establish measures to indicate, by the use of markings such as stamps, tags, labels, routing cards, or other suitable i

means, the status of inspections and tests performed upon individual items of the packaging. These measures must provide for the identification of items which have satisfactorily passed required inspections and tests where necessary to preclude inadvertent bypassing of the inspections and tests, j

106

[7590-01)

(b) The licensee shall establish measures to identify the operating status of components of the packaging, such as tagging valves and switches, to prevent inadvertent operation.

S 71.131 Nonconforming materials, parts, or components.

The licensee shall establish measures to control materials, parts, or components which do not conform to the licensee's requirements in order to prevent their inadvertent use or installation. These measures must include, as appropriate, procedures for identification, documentation, segregation, disposition, and notification to affected organizations. Nonconforming items must be reviewed and accepted, rejected, repaired, or reworked in accordance with documented procedures.

S 71.133 Corrective action.

The licensee shall establish measures to assure that conditions ad-verse to quality, such as deficiencies, deviations, defective material and equipment, and nonconformances, are promptly identified and cor-rected. In the case of a significant condition adverse to quality, the measures must assure that the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condition, and the corrective action taken must be documented and reported to appropriate levels of management.

S 71.135 Quality assurance records.

(a) The licensee shall maintain sufficient written records to  !

furnish evidence of activities affecting quality. The records must j 107  :

[7590-01) include the following: design records, records of use and the results of reviews, inspections, tests, audits, monitoring of work performance, and materials analyses. The records must include closely related data such as qualifications of personnel, procedures, and equipment. Inspection and test records must, at a minimum, identify the inspector or data recorder, the type of observation, the results, the acceptability, and the action taken in connection with any deficiencies noted. Records must be identifiable and retrievable. Records pertaining to the fabrication of the package must be retained for the life of the package to which they apply. Records pertaining to the use of the package for shipment of radioactive material must be retained for a period of 3 years after the shipment.

(b) The licensee shall establish a records retention program which is consistent with the applicable regulations, designating factors such as duration, location, and assigned responsibility.

S 71.137 Audits.

The licensee shall carry out a comprehensive system of planned and periodic audits to verify compliance with all aspects of the quality assurance program and to determine the effectiveness of the program. The audits met te performed in accordance with written procedures or check-lists by appropriately trained personnel not having direct responsibil-ities in the areas being audited. Audited results must be documented and reviewed by management having responsibility in the area sudited.

Follow-up action, including reaudit of deficient areas, must be taken where indicated.

)

108

- .i

[7590-01]

I Appendix A - Determination of At and A2 I. Values of As and A 2 for individual radionuclides, which are the bases for many activity limits elsewhere in these regulations are given in Table A-1. The curie values specified are obtained by extending the Terabecquerel (TBq) figure, after conversion to Curies, (Ci) to 3 significant figures. This ensures that the magnitude of As and A2 in Ci is always equal to that in TBq to a tenth of 1 percent.

II. For individual radionuclides whose identities are known, but which are not listed in Table A-1, the determination of the values of As ana A2 shall require Commission approval, except that the values of At and A2 in Table A-2 may be used without obtaining Commission approval.

i III. In the calculations of A and t A2 for a radionuclide not in Table A-1, a single ladioactive decay chain in which radienuclides are present in their naturally occurring proportions and in which no daughter nuclide has a half-life either longer than 10 days or longer than that of the parent nuclide shall be considered as a single radio-nuclide, and the activity to be taken into account and the At or As value to be applied shall be those corresponding to the parent nuclide of that chain. In the case of radioactive decay chains ia which any daughter nuclide has a half-life either longer than 10 days or greater than that of the parent nuclide, the parent and those daughter nuclides shall be considered as mixtures of different nuclides.

109

IV. For mixtures of radionuclides whose identities and respective -

activities are known, the following conditions shall apply:

(a) For spacial form radioactive material:

I less than or equal to 1 (b) For other forms of radioactive material I

h less than or equal to 1 where B(i) is the activity of radionuclide i and Ag(i) and A2 (i) are the At and A2 values for radionuclide i, respectively.

Alternatively, an A2 value for mixtures may be determined as follows:

I A2 for mixture =

f )

where f(i) is the fraction of activity of nuclide i in the mixture and A (i) is the appropriate A2 value for nuclide i.

l i

1 110 t

i

oL w_ _-4 y _,-_6__ - _ _ _ A - 4 - LA ,.. -%

, (7590-01) i V. When the identity of each radionuclide is known but the individual 4

activities of some of the radionuclides are not known, the radio-nuclides may be grouped and the lowest As or A2 value, as appro-priate, for the radionuclides in each group may be used in applying -

the formulas in paragraph IV. Groups may be based on the total i

alpha activity and the total beta / gamma activity when these are known, using the lowest At or A2 values for the alpha emitters or beta / gamma emitters, respectively. l I

e i

1 ,

I i

1 l l

1  !

' l i

l i

r I

i i I i  !

i i

i 111 8

D

-~

i 1

Table A-1 An and A2 Values for Radionuclides ,

1 4

Specific Symbol of Element and An A A: A2 Activity I

Radionuclide Atomic Number (IBq) (Ci) (IBq) (Ci) (Ci/g) i 22 sac Actinium (89) 0.6 16.2 1x10-2 0.270 5.8x104 227Ac 40 1080 2x10-5 5.41x10-4 7.2x108 22 sac 0.6 16.2 0.4 10.8 2.2x108 805 2 54.1 2 54.1 3.1x104 10sgg Silver (47) ,

Ag 0.6 16.2 0.6.2 16.2 6.6x102

] 810"Ag 0.4 10.8 0.4 10.8 4.7x103 j 888Ag 0.6 16.2 0.5 13.5 1.6x105 2 cal Aluminum (13) 0.4 10.8 0.4 10.8 1.9x10-2 24:ge Americium (95) 2 54.1 2x10-4 5.41x10-3 3.2 j 242 Am 2 54.1 2x10-4 5.41x10-3 8.1x105 i  ;; 243.h 2 54.1 2x10-4 5.41x10-3 1.9x10-8

} 37Ar Argon (18) 40 1080 40 1080 1.0x105 i 39Ar 20 541 20 541 3.4x101

! 41Ar 0.6 16.2 0.6 16.2 4.3x107 j 42 ar 0.2 5.41 0.2 5.41 2.6x102 72As Arsenic (33) 0.2 5.41 0.2 5.41 1.7x108 l 73As 40 1080 40 1080 2.4x104

) 74As 1 27.0 0. 5 13.5 1.0x105 76As 0.2 5.41 0.2 5.41 1.6x108 ,

77As 20 541 0.5 13.5 1.1x10'

2

At Astatine (85) 30 811 2 54.1 2.1x108 l 893Au Gold (79) 6 162 6 162 9.3x105 l 194Au 1 27.0 1 27.0 4.1x105

! 195Au 10 270 10 270 3.7x103 4

89'Au 2 54.1 2 54.1 1.2x105 ,

M 8

a ,

C

_ - - - _ . _ _ . _ - - _ _ _ ___ _ _. _ _ _ _ _ - _ _ _ _ _ =_ _ _ _ _ . _ _ _ _ _ _.

~

i 1 Table A-1 (Cont.)

Specific

! Symbol of Element and A A A2 A2 Activity l Radionuclide Atomic Number (18q) (Ci) (TBq) (Ci) (Ci/g) i assAu 3 81.1 0.5 13.5 2.5x105 199Au 10 270 0.9 24.3 2.1x105 j 831Ra Barium (56) 2 54.1 2 54.0 8.7x104 d 133"Ba 10 270 0.9 24.3 6.1x105 j 8338a 3 81.1 3 81.1 4.0x102 i 340Ba 0.4 10.8 0.4 10.8 7.3x104

{ 7Be 8ery11ium(4) 20 541 20 540 3.5x105

} 808e 20 541 0.5 13.5 1.4x10-2 3 20sBi Bismuth (83) 0.6 16.2 0.6 16.2 4.2x104 20s8i 0.3 8.11 0.3 8.11 9.9x104 l 3 207 i 0.7 18.9 0.7 18.9 2.2x102

zio i 0.3 8.11 3x10 2 0.811 6.5x10 4

( 2:08i 0.6 16.2 0.5 13.5 1.2x105 2:28i 0.3 8.11 0.3 8.11 1.5x107 l 2478k Berkelium (97) 2 54.0 2x10-4 5.41x10-3 1.0 ,

2498k 40 1080 8x10-2 2.16 1.8x103

] 76Br Bromine (35) 0.3 8.11 0.3 8.11 2.6x108 I

778r 3 81.1 3 81.1 7.1x105

s28r 0.4 108 0.4 10.8 1.1x108 1 3'C Carbon (6) 1 270 0.5 13.5 8.4x10s l

! 14C 40 1080 2 54.0 4.6 i

43Ca Calcium (20) 40 1080 40 1080 1.1x10-2 i 0.9 l 45Ca 40 1080 24.3 1.9x104 l 47Ca 0.9 24.3 0.5 13.5 5.9x105 j 109Cd Cadmium (48) 40 1080 1 27.0 2.6x102 i  ?

! 2 i

'i i

_ - - _ - - , . - - - - - - _ . . - - - --=- -, - . . - - - , - - - . , . - , , - - - - - - - = . - - - - , . -_ . - - , . = -- , _ - . _ . - - . _ , ,

Table A-1 (Cont.)

Specific Symbol of Eles?nt and A An A2 A2 Activity ,

Radionuclide Atomic Number (IBq) (Ci) (TBq) (Ci) (Ci/g) 883"Cd 20 540 9x10-2 2.43 2.3x102 885"Cd 0.3 8.11 0.3 8.11 2.6x104 88SCd 4 108 0.5 13.5 5.1x105 83*Ce Cerium (58) 6 162 6 162 6.5x103 848Ce 10 270 0.5 13.5 2.8x104 8 4 3'Ce 0.6 16.2 0.5 13.5 6.6x105 144Ce 0.2 5.41 0.2 5.41 3.2x103 i

. 24*Cf Californium (98) 30 811 3x10-3 8.11x10-2 1.5x103 249Cf 2 54.1 2x10-4 5.41x10-3 3.1 2soCf 5 135 5x10-4 1.35x10-2 1.3x102 l

• l g 2s Cf 2 54.0 2x10-4 5.41x10-3 1.8 l 2s2Cf 0.1 2.70 1x10-3 2.70x10-2 6.5x102 i

2ssCf 40 1080 6x10-2 1.62 2.9x104 2s4Cf 3x10-3 8.11x10-1 6x10-4 1.62x10-2 8.5x103 3'C1 Chlorine (17) 20 541 0.5 13.5 3.2x10-2 l

ssC1 0.2 5.41 0.2 5.41 1.3x10s 240Cm Curium (%) 40 1080 2x10-2 0.541 2.0x104 24:Cm 2 54.1 0.9 24.3 1.5x104 242Cm 40 1080 1x10-2 0.270 3.3x103 l 243Cm 3 81.1 3210-4 8.11x10-3 4.2x108 244Cm 4 1080 4x10-4 1.08x10-2 8.2x105 24sCm 2 54.1 2x10-4 5.41x10-3 1.0x10-8 24sCm 2 54.1 2x10-4 5.41x10-3 3.1x10-8 247Ce 2 54.1 2x10-4 5.41x10-3 9.1x10-5 24sCm 4x10-2 1.08 5x10-5 1.35x10-3 3.1x10-3 -

l u".

l 55C0 Cobalt (27) 0.5 13.5 0.5 13.5 3.1x108 ;g 5'Co- 0.3 8.11 0.3 8.11 3.Ox104 L, s7 o 8.5x103 8 216 8 216 C 58 Co 40 1080 40 1080 5.9x108 58C0 1 27.0 1 27.0 3.1x104

Table A-1 (Cont.)

Specific Symbol of Element and A A A2 A2 Activity Radionuclide Atomic Number (IBq) (Ci) (TBq) (Ci) (Ci/g)

'OCo 0.4 10.8 0.4 10.8 1.1x103 58Cr Chromium (24) 30 811 30 811 9.2x104 329Cs Cesium (55) 4 108 4 108 7.6x105 338Cs 40 1080 40 1080 1.0x105

2Cs 1 27.0 1 27.0 1.5x105 13*"Cs 40 1030 9 243 7.4x108 83*Cs 0.6 16.2 0.5 13.5 1.2x103 835Cs 40 1080 0.9 24.3 8.8x10-4 88'Cs 0.5 13.5 0.5 13.5 7.4x104 137Cs 2 54.0 0.5 13.5 8.7x108 C

8*Cu Copper (29) 5 135 0.9 24.3 3.8x108 67Cu 9 243 0.9 24.3 7.9x105 8590y Dysprosium (66) 20 541 20 541 5.7x103 8'5 Dy 0.6 . 16.2 0.5 13.5 8.2x108 88'Dy 0.3 8.11 0.3 8.11 2.3x105 88SEr Erbium (68) 40 1989 0.9 24.3 8.2x104 17 Er 0.6 16.2 0.5 13.5 2.4x108 147EU Europium (63) 2 54.1 2 54.0 4.1x104 i4sEu 0.5 13.5 0.5 13.5 6.1x104 8**Eu 20 541 20 540 8.3x103 858[u 0.7 18.9 0.7 18.9 1.7x108 is2 Eu 0.6 16.2 0.5 13.5 2.2x108 is2Eu 0.9 24.3 0.9 24.3 1.9x102 854Eu 0.8 21.6 0.5 13.5 1.5x102 855Eu 20 541 2 54.1 1.4x103 r,

~

85'Eu 0.6 16.2 0.5 13.5 5.4x104 hh 88F Fluorine (9) 1 27.0 0.5 13.5 9.3x107 f, s2Fe Iron (26) 0.2 5.40 0.2 5.40 7.3x108 -

55Fe 40 1080 40 1080 2.2x102 59Fe 0.8 21.6 0.8 21.6 4.9x104

. = - _ - . -_ _ ._ . _ - . _ - _ - _ _ _ -. - _ -__ - _ . _ _ . - _ _

Table A-1 (Cont.)

l Specific l Symbol of Elemer,t and A A A2 A2 Activity Radionuclide Atomic Number (18q) (Ci) (IBq) (Ci) (Ci/g)

'O 40 1080 0.2 5.41 2.0x10-2 i fe l '7Ga Ga111tm(31) 6 157 6 162 6.0x105 l '8Ga 0.3 8.11 0.3 8.11 4.0x107 72Ga 0.4 10.8 0.4 10.8 3.lx108 l

846Gd Gadolinium (64) 0.4 10.8 0.4 10.8 1.8x104 l

84"Gd 3 81.1 3x10 4 8.11x10 3 2.9x108 853Gd 10 270 5 135 3.6x103 1

853Gd 4 108 0.5 13.5 1.1x108 0.3

'8Ge Germanium (32) 0.3 8.11 8.11 7.0x103 73Ge 40 1080 40 1080 1.6x105 77Ge 0.3 8.11 0.3 8.11 3.6x10' 3:! Hydrogen (1) See T-Tritium 172Hf Hafnium (72) 0.5 13.5 0.3 8.11 4.2x102 175Hf 3 81.1 3 81.1 1.1x104 888Hf 2 54.1 0.9 24.3 1.6x104 is2Hf 4 108 3x10-2 0.811 2.2x10-4 894gg Mercury (80) 1 27.0 1 27.0 9.7x102 895 Hg 5 135 5 135 4.0x105 197"Hg 10 270 0.9 24.3 6.6x105 197 Hg 10 270 10 270 2.5x105 203 Hg 4 108 0.9 24.3 1.4x104 8'3 40 1080 40 1080 2.2 Holmium (67) l 8 " "110 00 0.6 16.2 0.3 8.11 1. 8 368Ho 0.3 8.11 0.3 8.11 6.9x105 n23I Iodine (53) 6 162 6 162 1.9x10' -

l M 8

I a

! C 1

1

Table A-1 (Cont.)

Specific Symbc1 of Element and A A A2 A2 Activity Radionuclide Atomic Number (TBq) (Ci) (IBq) (Ci) (Ci/g) 32*I 0.9 24.3 0.9 24.3 2.5x105 32sl 20 541 2 54.1 1.7x104 82'I 2 54.1 0.9 24.3 7.8x104 3291 Unlimited Unlimited 1.Gx10-4 8311 3 81.1 0.5 13.5 1.2x105 3321 0.4 10.8 0.4 10.8 1.1x107 8331 0.6 16.2 0.5 13.5 1.1x10' 8341 0.3 8.11 0.3 8.11 2.7x107 8351 0.6 16.2 0.5 13.5 3.5x10' 888In Indium (49) 2 54.1 2 54. 1 4.2x105 0- 883"In 4 108 4 108 1.6x107 03 884"In 0.3 8.11 0.3 8.11 2.3x104 885"In 6 162 0.9 24.3 6.1x10' 88'Ir Iridium (77) 10 270 10 270 5.2x104 8*8Ir 0.7 18.9 0.7 18.9 6.2x104 392fr 1 27.0 0.5 13.5 9.1x103 883 Ir 10 270 10 270 5.7x104 884Ir 0.2 5.41 0.2 5.41 8.5x105 42K Potassium (19) 0.2 5.41 0.2 5.41 6.0x108 43K 1.0 27.0 0.5 13.5 3.3x108 a:Kr Krypton (36) 40 1080 40 1080 2.1x10-2 asKr" 6 162 6 162 8.4x108 asKr 20 541 10 270 4.0x102 87Kr 0.2 5.41 0.2 5.41 2.8x107 337ta Lanthanum (57) 40 1080 2 54.1 4.4x10-2 ,_

U 8

a C

Table A-1 (Cont.)

Specific Symbol of Element and Ai A A2 A2 Activity I Radionuclide Atomic Number (TBq) (CI) (TBq) (Ci) (Ci/g) 840La 0.4 10.8 0.4 10.8 5.6x105 LSA low Specific Activity Material Definition (See S 71.4) 172Lu Lutetium (71) 0.5 13.5 0.5 13.5 4.6x104 173 a 8 216 8 216 1.5x103 174 Lu 20 541 8 216 5.4x103 174Lu 8 216 4 108 5.7x102 177tu 30 811 0.9 24.3 1.1x105 MFP Mixed Fission Products (use formula for mixture or Table A-2) 2aMg Magnesium (12) 0.2 5.41 0.2 5.41 5.2x10' 52Mn Manganese (25) 0.3 8.11 0.3 8.11 4.4x105

e. 53Mn Unlimited Unlimited 3.6x10-3 as 54Mn 1 270 1 27.0 8.3x103 5'Mn 0.2 5.41 0.2 5.41 2.2x107 83Mo Molybdenum (42) 40 1080 7 189 38 SSMo 0.6 16.2 0.5 13.5 4.7x105 83N Nitrogen (7) 0.6 16.2 0.5 13.5 1.5x109 22Na Sodium (11) 0.5 13.5 0.5 13.5 6.3x103 24ga 0.2 5.41 0.2 5.41 8.7x108 92 Mb Niobium (41) 0.7 18.9 0.7 18.9 1.4x105 93"Nb 40 1080 6 162 1.1x103 84Nb 0.6 16.2 0.6 16.2 1.9x10-1 85Nb 1 27.0 1 27.0 3.9x104 97Nb 0.6 16.2 0.5 13.5 2.6x107 847Nd Neodymium (60) 4 108 0.5 13.5 8.0x104 848Nd 0.6 16.2 0.5 13.5 1.1x107 ,,

3 l8 6

C

^

l Table A-1 (Cont.)

Specific Symbol of Element and A A A2 A2 Activity '

Radionuclide Atomic Number (TBq) (Ci) (IBq) (Ci) (Ci/g)

SSNi Nickel (28) 40 1080 40 1080 8.11x10-2

'3Ni 40 1080 30 811 4.6x108

'5Ni 0.3 8.11 0.3 8.11 1.9x107 23sNp Neptunium (93) 40 1080 40 1080 1.4x103 2ssNp 7 189 1x10-3 2.70x10-2 6.0x105 237'Np 2 54.1 0.5 13.5 6.9x10-4 239 Np 6 162 0.5 13.5 2.3x105 sasgs Ossium (76) 1 27.0 1 27.0 7.3. -d-3 888 Os 40 1080 40 1000 1.2x10' 8830s 10 270 0.9 24.3 4. Gxit24

.- 8930s 0.6 16.2 0.5 13.5 5.3x105 0; 8940s 0.2 5.41 0.2 5.41 1.1x102 s2P Phosphorus (15) 0.3 8.11 0.3 8.11 2.9x105 ,

23P 40 1080 0.9 24.3 1.6x105 '

2soPa Protactinium (91)2 54.1 0.1 2.70 3.2x104 238Pa C.6 16.2 6x10-5 1.62x10-3 4.5x10-2 233Pa 5 135 0.9 24.3 2.1x104 20:Pb Lead (82) 1 27.0 1 27.0 1.7x10' ,

202Pb 40 1080 2 54.1 5.9x20-3 - ,

203Pb 3 81.1 3 81.1 3.0x105 20sPb Unlimited Unlimited 5.8x10-5 2 oPb 0.6 16.2 9x10-3 0.243 8.8x108 2:2Pb 0.3 8.11 0. 3 8.11 1.4x106 i

~

sosPd Pa11adium(46) 40 1980 40 1080 7.5x104 307Pd Unlimited Unlimited 4.8x10-4 -,

3 8

a

~ . . _ _ - . _

- l Table A-1 (Cont.)

Specific Syd>ol of Element and An A A2 A2 Activity Radionuclide Atomic Number (TBq) (Ci) (IE<1) (Ci) (Ci/g) 809Pd 0.6 16.2 0.5 13.5 2.1x10-4 843Pm PrometP.ium(61) 3 81.1 3 81.1 3.4x103 144Pm 0.6 16.2 0.6 16.2 2.6x100 84SPs 30 811 7 189 1.4x102 j 8'7Pm 40 1080 0.9 24.3 9.4x132 l 84*"Pm 0.5 13.5 0.5 13.5 2.1x104 84*Pm 0.6 81.1 0.5 13.5 7.5x105 l 158Pe 3 16.2 0.5 13.5 4.2x105 20sPo Polonium (84)3 40 1080 2x10-2 0.541 5.9x102 209Po 40 1000 2x10-2 0.541 1.7x108 g 21oPo 40 1080 2x10-2 0.541 5.41x103 o 342Pr Praseodymium (59)0.2 5.41 0.2 5.41 1.2x104 843Pr 4 108 0.5 13.5 6.5x104 ,

assPt Platinum (78) 0.6 16.2 0.6 16.2 6.8x104  !

191Pt 3 81.1 3 81.1 2.3x105 l

883"Pt 40 1080 9 243 2.0x105 3.7 193gt 40 108G 40 1000 895 Pt 10 270 2 54.1 1.6x105 197"Pt 10 270 0.9 24.3 1.2x107 ~

197Pt 20 541 0.5 13.5 8.8x105 22 cpu Plutonium (94) 7 189 7x10-4 1.89x10-2 5.3x102 237Pu 20 541 20 541 1.2x104 1 2ssPu 2 54.1 2x10-4 5.41x10-3 1.7x108 l 23*Pu 2 54.1 2x10-4 5.41x10-3 6.2x10-2 2<oPu 2 54.1 2x10-4 5.41x10-3 2.3x10-8 m u

0 l

..- _ _ _ _ _____... _ _.___. _ _ __.____ _p _ _ _

1 _

}

l' Table A-1 (Cont.)

I

! Specific

Symbol of Element and A A Ar A2 Activity 1 Radionuclide Atomic Number (IBq) (Ci) (TBq) (Ci) (Ci/g) i l

24 Pu 40 1000 1x10-2 0.270 1.1x102 7{ 242Pu 2 54.1 2x10-4 5.41x10-3 3.9x10-3 24*Pa 0.3 8.11 2210-4 5.41x10-2 1.9x10-5 22sRa Radium (86) 0.6 16.2 3x10-2 0.811 5.0x104 22*Ra 0.3 8.11 6210-2 1.62 1.6x105 22sRa 0.6 16.7 2x10-2 0.541 3.9x10-4 1

22sRa 0.3 8.11 2x10-2 0.541 1.0 22sRa 0.6 16.2 4x10-2 1.08 2.3x102 88Rb Reidium(37) 2 54.1 0.9 24.3 8.2x10-8 ssRb 2 54.1 2 54.1 1.9x104 1

U

• • Rb 1 27.0 0.9 24.3 4.7x104 l ssRb 0.3 S.11 0.3 8.11 8.11x104 i s7Rb Unlimited Unlimited 6.6x10-s t Rb (natural) Unlimited Unlimited 1.8x10-s

{ a3Re Rheninn(75) 5 135 5 135 1.0x103 i

I 8***Re 3 81.1 3 81.1 4.2x103 i as*Re 1 27.0 1 27.0 1.9x104

! as'Re 4 108 0.5 13.5 1.9x105 as7Re Unlimited unlimited 3.8x10-s ssRe 0.2 5.41 0.2 5.11 1.0x10' is9Re 4 108 0.5 13.5 4.9x103 Re (natural) Unlimited Unlimited 2.4x10-s 1 **Rh Rhodium (45) 2 54.1 2 54.1 6.7x105

! 888 h 4 108 4 108 1.2x103 202 2 54.1 0.9 24.3 1.2x108 8

i e i O i

Table A-1 (Cont.)

Specific Symbol of Element and A An A2 Activity Radionuclide Atomic Number (TBq) (Ci) (TBq) ,oi) (Ci/g) to2gh 0.5 13.5 0.5 15.d 6.2 103 Rh 40 1080 40 1080 3.2x10 7 105Rh 10 270 0.9 24.3 8.2x105 222Rn Radon (86) 0.2 5.41 4x10-3 0.108 1.5x105 87Ru Rutheniam(44) 4 108 4 108 5.5x105 103Ru 2 54.1 0.9 24.3 3.2x104 105Ru 0.6 16.2 0.5 13.5 6.6x108 106Ru 0.2 5.41 0.2 5.41 3.4x103 355 Sulfur (16) 40 ~080 t 2 54.1 4.3x104 122Sb Antincis(51) 0.3 8.11 0.3 8.11 3.9x105

324Sb 0.6 16.2 0.5 13.5 1.8x104 n2 x2sSb 2 54.1 0.9 24.3 1.4x103 x2sSb 0.4 10.8 0.4 10.8 8.3x104 44Sc Scandium (21) 0.5 13.5 0.5 13.5 1.8x107 48Sc 0.5 13.5 0.5 13.5 3.4x104 475c 9 243 0.9 24.3 8.2x105 48Sc 0.3 8.11 0.3 8.11 1.5x108 SCO Surface Contaminated Object Definition (See S 71.4) 1 75Se Salenium(34) 3 81.1 3 81.1 1.4x104 795e 40 1060 2 54.1 7.0x10-2 31Si Silicon (14) 0.6 16.2 0.5 13.5 3.9x107 323; 40 1080 0.2 5.41 1.7x101 l

145Se Samarium (62) 20 541 20 541 2.6x103 147Se Unlimited Unlimited 2.0x10-8 151Sm 40 1080 4 108 2.6x101

b -

l 1

j Table A-1 (Cont.)

l ..

] Specific

! Symbol of Element and An Ai A2 A2 Activity l Radionuclide Atomic Number (TBq) (Ci's (TBq) (Ci) (Ci/g) 1535m 4 108 0.5 13.5 4.4x105 113gn Tin (50) 4 108 4 108 1.0x104 117 6 162 2 54.1 8.0x104 189"Sn Sn 40 1080 40 1080 4.4x103 22tmSn 40 1080 0.9 24.3 3.9x101 22sSn 0.6 16.2 0.5 13.5 8.5x103 12sSn 0.2 5.41 0.2 5.41 1.1x105

12sSn 0.3 8.11 0.3 8.11 2.8x10-2

s2gr Strontium (38) 0.2 5.41 0.2 5.41 6.4x104 85 Sr 5 135 5 135 3.2x107 g 8sgr 2 54.1 2 54.1 2.4x104 W

87 Sr 3 81.1 3 81.1 1.2x107 l 895r 0.6 16.2 0.5 13.5 2.9x104

! 90Sr 0.2 5.41 0.1 2.70 1.5x102 j 91Sr 0.3 8.11 0.3 8.11 3.6x108 925r 0.2 5.41 0. 1.3x107 1080 40{1) 5.41(1) 1080 9.7x103 T Tritium (1) 40 178Ta Tantalum (73) 1 27.0 1 27.0 1.1x10s 179Ta 30 811 30 811 1.2x103 182Ta 0.8 21.6 0.5 13.5 6.2x10a 1s7Tb Terbium (65) 40 1080 10.8 T O 1.5x101 isSTb 1 27.0 0.7 16.9 1.9 180Tb 0.9 24.3 0.5 13.5 1.1x104 95"Tc Technetium (43) 2 54.1 2 54.1 2.2x104 96"Tc 0.4 10.8 0.4 10.8 3.8x107 II)Also, for liquids only, a concentration limit of not greater than 27.0 Ci/t (1 TBq/E)

)

Table A-1 (Cont.)

Specific Symbol of Element and An An A2 A2 Activity Radionuclide Atomic Number (TBq) (Ci) (TBq) (Ci) (Ci/g) 0.4 10.8 0.4 10.8 3.2x105 8'[c 97 Tc 40 1080 40 1080 1.5x104 97Tc Unlimited Unlimited 1.4x10-3 98 c 0.7 18.9 0.7 18.9 2.4x10-3 99 Tc 8 216 8 216 5.2x108 99T'c 40 10C0 0.9 24.3 1.7x10-2 tis e Te11urium(52) 0.2 5.41 0.2 5.41 1.8x105 12i Te 5 135 5 135 7.0x103 2 54.1 2 54.1 6.3x104 121[e 189 7 189 9.1x103 123 Te 7

125"Te 30 811 9 243 1.8x104 l 127"Te 20 541 0.5 13.5 4.0x104 127 e 20 541 0.5 13.5 2.6x106 12 Te 0.6 16.2 0.5 13.5 2.5x104 l 129Te 0.6 16.2 0.5 13.5 2.0x107 131"Te 0.7 18.9 0.5 13.5 8.0x105 is2Te 0.4 10.8 0.4 10.8 3.1x105 227Th Thorium (90) 9 243 1x10-2 0.270 3.2x104 22sTh 0.3 8.11 4x10-4 1.08x10-2 8.3x102 8.11 8.11x10-4 2.1x10-1 229Th 0.3 3x10-5 230Th 2 54.1 2x10-4 5.41x10-3 1.9x10-2 l 231Th 40 1080 0.9 24.3 5.3x105 232Th Unlimited Unlimited 1.1x10 7 234Th 0.2 5.41 0.2 5.41 2.3x104 Th (natural) Unlimited Unlimited 2.2x10-7

'e l

l l

Table A-1 (Cont.)

Specific

Symbol of Element and A A A: A2 Activity I Radionuclide Atomic Number (TBq) (Ci) (IBq) (Ci) (Ci/g) 44Ti Titanium (22) 0.5 13.5 0.2 5.41 1.7x102 ,

i 200T1 Tha11ium(81.1) 0.8 21.6 0.8 21.6 5.8x105

20 T1 10.8 270 10.8 270 2.2x105 1 202T1 2 54.1 2 54.1 5.41x104

204T1 4 108 0.5 13.5 4. x102 l 167Tm Thulium (69) 7 189 7 189 8.11x104 j issTm 0.8 21.6 0.8 21.6 9.1x103 1

170Tm 4 108 0.5 13.5 6.0x103 171Tm 40 1080 10.8 270 1.1x103 2300 Uranium (92) 40 1080 1x10-2 0.270 2.7x104 3 2 232U 3 81.1 3x10-4 8.11x10-3 2.1x101 233U 10.8 270 1x10-3 2.70x10-2 9.5x10-3 234U 10.8 270 1x10-3 2.70x10-2 6.2x10-3 23s0 Unlimited Unlimited 2.1x10-6 2380 10.8 270 1x10-3 2.70x10-2 6.3x10-5

. 23su Unlimited Unlimited 3.3x10-7 1

U (natural) Unlimited Unlimited 7.1x10-7 U (enriched 5% or less) Unlimited Unlimited (see Table A-3) l U (enriched more than 5%) 10.8 270 1x10-3 2.70x10-2 (see Table A-3)

U (depleted) Unlimited Unlimited (see Table A-3) i

, 4sv Vanadium (23) 0.3 8.11 0.3 8.11 1.7x105 49V 40 1080 40 1080 8.11x103 Tungsten (74) 27.0 27.0 3.4x104 178W 1 1

as1W 30 811 30 611 5.0x103 185W 40 1080 0.9 24.3 9.7x10-3 O

i i

^

i

] Table A-1 (Cont.)

l Specific I Symbol of Element and Ai A A2 A2 Activity

! Radionuclide Atomic Number (IBq) (Ci) (TBq) (Ci) (Ci/g) 187W 2 54.1 0.5 13.5 7.0x10-5 l tasW 0.2 5.41 0.2 5.41 1.0x104 122Xe Xenon (54) 0.2 5.41 0.2 5.41 1.3x108 223Xe 0.2 5.41 0.2 5.41 1.2x107 127Xe 4 108 4 108 2.8x104 131"Xe 40 1080 40 1080 1.0x105 133Xe 20 541 20 541 1.9x105 1^5Xe 4 108 4 108 2.5x105 I. 8*Y Yttrium (39) 2 54.1 2 54.1 4.5x101 4

88Y 0.4 10.8 0.4 10.8 1.4x104 5 S0 1 0.2 5.41 0.2 5.41 2.5x105

, 91"Y 2 54.1 2 54.1 4.lx107 i 91Y 0.3 8.11 0.3 8.11 2.5x104 4 92Y 0.2 5.41 0.2 5.41 9.5x108 S3Y 0.2 5.41 0.2 5.41 3.2x108 188Yb Ytterbius(70) 3 81.1 3 81.1 2.3x105 175Yb 30 811 0.9 24.3 1.8x105 5 2 54.1 2 54.1 8.0x103 Zinc (30) l 8Jn 8 Zn 2 54.1 0.5 13.5 3.3x108

! 892n 4 108 0.5 13.5 5.3x107 1

asZr 3 81.1 3 81.1 1.7x104

' Zirconium (40)

S3Zr 40 1080 0.2 5.41 3.5x10 3 .

95Zr 1 27.0 0.9 24.3 2.1x104 97Zr 0.3 8.'11 0.3 8.11 2.0x108 y

40 i  ?

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[7590-01]

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Table A-2 General Values for At and A2 l At A2 Contents (TBq) (Ci) (TBq) (Ci)

Only beta or gamma emitting 0.2 5 0.02 0.5 i nuclides are known to be present Alpha emitting nuclides are known 0.10 2.70 2 x 10 5 5.41 x 10 4 i.o be present or no relevant data are available Table A-3 Activity-rnass Relationships for Uranium / Thorium Thorium and Specific Activity Uranium Enrichment

• wt % 23su present Ci/g g/Ci 0.45 5.0 x 10 7 2.0 x 108 0.72 7.06 x 10 7 1.42 x 108

1. 0 7.6 x 10 7 1.3 x 108

1. 5 1.0 x 10 8 1.0 x 108 5.0 2.7 x 10 8 3.7 x 105 10.0 4.8 x 10 8 2.1 x 105 20.0 1.0 x 10 5 1.0 x 105 35.0 2.0 x 10 5 5.0 x 104 -

50.0 2.5 x 10 5 4.0 x 104 90.0 5.8 x 10 5 1.7 x 104 93.0 7.0 x 10 5 1.4 x 104 95.0 9.1 x 10 5 1.1 x 104 Natural Thorium 2.2 x 10 7 4.6 x 108 "The figures for uranium include representative values for the activity of the uranium-234 which is concentrated during the enrichment process. The activity for Thorium includes the equilibrium concentration of Thorium-228.

Dated at Rockville MD this 23rd day of May 1988.

fortheNuclearRegulatoryCommission.

//M' { ( fdt A Viftor Stello, Jr. C' Executive Direc)(r or Operations i

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127 l

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