ML23152A139

PR-050, 070 - 62FR63911 - Criticality Accident Requirements

ML23152A139
Person / Time
Issue date:	12/03/1997
From:	Callan L NRC/EDO
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References
PR-050, PR-070, 62FR63911
Download: ML23152A139 (1)

Text

ADAMS Template: SECY-067 DOCUMENT DATE: 12/03/1997 TITLE: PR-050, 070 - 62FR63911 - CRITICALITY ACCIDENT REQUIREMENTS CASE

REFERENCE:

PR-050, 070 62FR63911 KEYWORD: RULEMAKING COMMENTS Document Sensitivity: Non-sensitive - SUNSI Review Complete

STATUS OF ROLEMAICING PROPOSED ROLE: PR-050, 070 OPEN ITEM (Y/N) N ROLE NAME: CRITICALITY ACCIDENT REQUIREMENTS PROPOSED ROLE FED REG CITE: 62FR63911 PROPOSED ROLE PUBLICATION DATE: 12/03/97 NUMBER OF COMMENTS: 9 ORIGINAL DATE FOR COMMENTS: 01/02/98 EXTENSION DATE: I I FINAL ROLE FED. REG. CITE: 63FR63127 FINAL ROLE PUBLICATION DATE: 11/12/98 NOTES ON: DIRECT FINAL ROLE ALSO PUB. ON 12/03/97 - SEE 62FR63825). DFR WIT STATUS HDRAWN (SEE 63FR9402, PUBLISHED ON 2/25/98). FINAL ROLE /S/'D BY OF ROLE : EDO HISTORY OF THE ROLE PART AFFECTED: PR-050, 070 ROLE TITLE: CRITICALITY ACCIDENT REQUIREMENTS PROPOSED ROLE PROPOSED RULE DATE PROPOSED ROLE SECY PAPER: 97-155 SRM DATE: 08/19/97 SIGNED BY SECRETARY: 11/14/97 FINAL ROLE FINAL ROLE DATE FINAL ROLE SECY PAPER: SRM DATE: I I SIGNED BY SECRETARY: 10/28/98 STAFF CONTACTS ON THE ROLE CONTACTl: STAN TUREL MAIL STOP: T-9F31 PHONE: 415-6234 CONTACT2 : MAIL STOP: PHONE:

DOCKET NO. PR-050, 070 (62FR639ll)

In the Matter of CRITICALITY ACCIDENT REQUIREMENTS DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT 12/09/97 11/14/97 FEDERAL REGISTER NOTICE - PROPOSED RULE 12/29/97 12/22/97 COMMENT OF COMMONWEALTH EDISON COMPANY (THOMAS J. KOVACH, VICE PRESIDENT) ( 1) 12/29/97 12/24/97 COMMENT OF CAROLINA POWER &LIGHT COMPANY (D. B. ALEXANDER) ( 2) 01/05/98 12/31/97 COMMENT OF SOUTHERN NUCLEAR OPERATING COMPANY, INC.

(H. L. SUMNER, JR., VICE PRESIDENT) ( 3) 01/05/98 01/02/98 COMMENT OF NUCLEAR ENERGY INSTITUTE (DAVID J. MODEEN, DIRECTOR) ( 4) 01/06/98 01/02/98 COMMENT OF NORTHERN STATES POWER COMPANY (MARCUS H. VOTH) ( 5)

. 1/09/98 01/08/98 COMMENT OF LINDA R. DEWHIRST ( 6) 01/09/98 01/02/98 COMMENT OF DETROIT EDISON (NORMAN K. PETERSON, DIRECTOR) ( 7) 01/14/98 01/07/98 COMMENT OF PECO ENERGY COMPANY (G. A. HUNGER, JR., DIR. LICENSING) ( 8) 01/15/98 01/13/98 COMMENT OF CARL STEPHENSON ( 9) 02/20/98 02/20/98 FEDERAL REGISTER NOTICE: DIRECT FINAL RULE; WITHDRAWAL 11/13/98 10/28/98 FEDERAL REGISTER NOTICE - FINAL RULE

50.J-1o DOCKE TED

(~~FR&3&JII) [7590-01 -PJ USNRC

°98 NOV 13 A8 :20 NUCLEAR REGULA TORY COMMISSION OH-I.

10 CFR Parts 50 and 70 RLL,.

AOJlJ[J1 TF RIN: 3150-AF87 Criticality Accident Requirements AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) is amending its regulations to give licensees of light-water nuclear power reactors greater flexibility in meeting the requirement that licensees authorized to possess more than a small amount of special nuclear material (SNM) maintain a criticality monitoring system in each area in which the material is handled, used, or stored. This action is taken as a result of the experience gained in processing and evaluating a number of exemption requests from such licensees and NRC's safety assessments in response to these requests that concluded that the likelihood of criticality was negligible.

~ ! L f,/9 98 EFFECTIVE DATE: The final rule is effective on ... ~ao da,s after i,ublicatio11 i11 tl,e Fede,al P~- ~ ,,;, 2 / qg-a::r. ~?>FRlo3I ::Z"/

FOR FURTHER INFORMATION CONTACT: Michael T. Jamgochian, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone:

{301) 415-3224; e-mail: mtj1@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Background The U.S. Nuclear Regulatory Commission (NRC) is amending its regulations to give persons licensed to construct or operate light-water nuclear power reactors the option of either

• meeting the criticality accident requirements of paragraph {a) through (c) of 10 CFR 70.24 in handling and storage areas for SNM, or electing to comply with certain requirements that are set forth in a new Section 50.68 in 1o CFR Part 50. The requirements in Section 50.68 are generally the requirements that the NRC has used to grant specific exemptions from the requirements of 10 CFR 70.24. In addition, the NRC is deleting the current text of Section 70.24(d) concerning the granting of specific exemptions from Section 70.24 because it is redundant to 10 CFR 70.14{a). Section 70.24(d) is rewritten to provide that the requirements in paragraphs {a) through {c) of 10 CFR 70.24 do not apply to holders of a construction permit or operating license for a nuclear power reactor issued under 1o CFR Part 50, or combined licenses issued under 10 CFR Part 52, if the holders comply with the requirements of 10 CFR 50.68{b).

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11. Discussion On December 3, 1997 (62 FR 63825), the NRC published a direct final rule in the Federal Register that would have provided persons licensed to construct or operate light-water nuclear power reactors with the option of either meeting the criticality accident requirements of paragraph (a) of 10 CFR 70.24 in handling and storage areas for SNM, or electing to comply with requirements that would be incorporated into 10 CFR Part 50 at 10 CFR 50.68. A direct final rule (62 FR 63825) and a parallel proposed rule (62 FR 63911) amending Parts 70 and 50 were published in the Federal Register on December 3, 1997. The statement of considerations for the direct final rule and the proposed rule stated that if significant adverse comments were received on the direct final rule, the NRC would withdraw the direct final rule and would address the comments in a subsequent final rule. Significant adverse comments :were received from the public, and on February 25, 1998, the NRC published a notice withdrawing the direct final rule and revoking the regulatory text. Since the direct final rule had an effective date of February 17, 1998, it was necessary for the February 25, 1998 notice to revoke the regulatory text which became effective on February 17, 1998, as well as to withdraw the direct final rule.

With the withdrawal and revocation, the *proposed rule is the only regulatory proposal remaining. The NRC has determined to modify the proposed rule to address public comments and to make several editorial clarifications. The analysis of and response to the public comments to the proposed rule are set forth below.

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111. Comments on the Proposed Rule The NRC received comments on the December 3, 1997, proposed rule (62 FR 63911) from Commonwealth Edison, Carolina Power & Light Company, Southern Nuclear Operating Company, Nuclear Energy Institute, Northern States Power Company, Trojan Nuclear Plant, and Detroit Edison. Copies of the letters are available for public inspection and copying for a fee at the Commission's Public Document Room, located at 2120 L Street, NW. (Lower Level),

Washington, DC. Many of the comment letters suggested editorial type changes, some of which have been incorporated into this final rule. The comments are classified into nine general comments and are addressed as follows:

Comment 1: The proposed rule should not prohibit licensees from applying for exemptions under the guidelines of 10 CFR 70.14 and should contain provisions to note that any existing approved exemptions remain valid.

Response: Even though the wording of paragraph (d) in the current version of 10 CFR 70.24, which provides for applying for exemptions should "good cause" exist, is being deleted,

• licensees are not prohibited from applying for such exemptions under the guidelines of

• paragraph (a) of ~O CFR 70.14, "Specific Exemptions."

The standard for issuance of exemptions under Section 70.14 is essentially the same as the "good cause" criterion in paragraph (d) of Section 70.24. Therefore, its removal from Section 70.24(d) will not change the standard for, or otherwise serve to limit the granting of, exemptions to Section 70.24.

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This rulemaking does not affect the status of exemptions to the requirements of Section 70.24 that were previously granted by the NRC. A licensee currently holding an exemption to Section 70.24 may continue operation under its existing exemption (including any applicable conditions imposed as part of the granting of the exemption) and its current programs and commitments without any further action. Alternatively, a licensee currently holding exemptions to Section 70.24 may elect to comply with the new alternative-provided under Section 50.68(b),

but if it does so, its exemption would be inapplicable and would not serve as a basis for avoiding compliance with the criteria listed in Section 50.68(b). A licensee whose exemption was issued as part of its operating license and whose exemption contained conditions imposed as part of the granting of the exemption, need not apply for a license amendment to delete the exemption conditions as a prerequisite for complying with Section 50.68(b).

Comment 2: For many BWRs, optimum moderation calculations are not performed for the fresh fuel storage racks because administrative controls are in place to preclude these conditions. In accordance with vendor recommendations, compensatory measures have been established to preclude an optimum moderation condition in the fresh fuel storage racks. The rule should contain a provision t~at exempts this requirement if adequate controls have been established to preclude an optimum moderation condition.

Response: The NRC agrees and has added the following provision to 10 CFR 50.68(b)(3): "This evaluation need not be performed if administrative control and/or design features prevent such moderation, or if fresh fuel storage racks are not used."

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Comment 3. The rule should eliminate the reference to General Design Criterion 63 (GDC 63) and should describe the underlying monitoring requirements.

Response: The reference to GDC 63 was initially incorporated to ensure that licensees receiving an exemption to 10 CFR 70.24 would not erroneously view the exemption as the basis for removing from the spent fuel pool area radiation monitors that were installed to meet other monitoring requirements, such as those contained in 10 CFR 20.1501 and GDC 63. This rule change does not affect these other monitoring requirements; therefore, referencing GDC 63 has been deleted.

Comment 4. Placing a limit on enrichment offers no direct safety benefit and should not be included.

Response: The NRC disagrees with the comment. The maximum allowable nominal enrichment of reactor fuel is currently limited to 5-weight percent on the basis of possible criticality concerns even in a dry environment, as well as currently approved extensions to 10 CFR 51.52 based on an environmental impact study for enrichments higher than 5-weight percent. Any future approved enrichment extension can be readily handled by modifying this criterion.

Comment 5. Replace "may not permit' with "shall prohibit the" in Criterion (1).

Response: The NRC agrees and has used the phrase suggested by the commenter.

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Comment 6. Use of apure water" and "unborated water" should be consistent.

Response: The NRC agrees. The final rule uses the term "unborated water."

Comment 7. Criteria (2) and (3) should not be applicable if the licensee does not use the fresh fuel storage racks.

Response: The NRC agrees and has added the following provision to 10 CFR 50.68(b)(2) and (b)(3): "This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used."

Comment 8. The meaning of "transportationD in criterion (1) is unclear.

Response: The NRC agrees and has deleted the term.

Comment 9. The phrase "maximum permissible U-235 enrichment" in Criteria (2), (3),

and (4) should be replaced by the phrase "maximum fuel assembly reactivity."

Response: The NRC agrees and has used the phrase suggested by the commenter.

IV. Section-by-Section Analysis 10 CFR Section 50.68 7

Paragraph (a) of Section 50.68 allows a nuclear power plant licensee (including a holder of either a construction permit or a combined operating license) the option of complying with Section 70.24(a) through (c), or complying with the requirements in paragraph (b) of Section 50.68. The corresponding provision in Section 70.24 is paragraph (d).

Paragraph (b) sets forth eight specific requirements which a licensee must comply with so long as it chooses under the provisions of Section 50.68 to avoid compliance with the requirements of Section 70.24(a) through (c).

~ licensee currently holding an exemption to Section 70.24 may elect to comply with the new alternative provided under Section 50.68, but if it does so, its exemption to Section 70.24 is inapplicable to, and would not serve as a basis for avoiding compliance with the eight criteria in Section 50.68(b).

10 CFR Section 70.24 Paragraph (d)(1) of Section 70.24 allows a nuclear power plant licensee (including a holder of either a construction pennit or a combined operating license) the option of complying with Section 70.24(a) through (c), or complying with th&requirements in* 10:CFR Section 50.68.

This paragraph is the corresponding provision to Section 50.68(a).

Paragraph (d)(2) clarifies that the status of exemptions to the requirements of Section 70.24 that were previously granted by the NRC continue unaffected by this rulemaking. A licensee currently holding an exemption to Section 70.24 may continue operation under its existing exemption (including any applicable conditions imposed as part of the grant of the exemption) and its current programs and commitments without any further action.

A license that seeks an exemption from the requirements of Section 70.24 must meet the criteria for an exemption under Section 70.14. The standard for issuance of exemptions 8

remains unchanged from the old rule, since the Commission regards the former "good cause*

criterion under the previous version of Section 70.24(d) as being essentially the same as the standard for issuance of exemptions under Paragraph 70.14.

V. Metric Policy On October 7, 1992, the Commission published its final Policy Statement on Metrication.

According to that policy, after January 7, 1993, all new regulations and major amendments to existing regulations were to be presented in dual units. The new addition and amendment to the regulations contain no units.

VI. Finding of No Significant Environmental Impact

• The NRC 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, would not be a major Federal action significantly affecting the qualify of the human environment; and therefore, an environmental impact statement is not required. The final rule provides an alternative to existing requirements on criticality monitoring. The alternative method contained in the final rule in the new Section 50.68 represents a codification of the criteria currently used by the NRC for granting exemptions from the criticality monitoring requirements in 10 CFR 70.24(a). These criteria provide an acceptable alternative for assuring that there are no inadvertent criticality events of special nuclear material at nuclear power reactors, which is the purpose of the criticality monitoring requirements in Section 70.24(a).

Experience over 15 years has demonstrated that the alternative criteria have been effective in 9

preventing Inadvertent-criticality events, and the NRC concludes that as a matter of regulatory efficiency, there is no purpose to requiring licensees to apply for and obtain exemptions from requirements of Section 70.24(a) if they adhere to the alternative criteria In the new Section 50.68. Since the alternative contained in Section 50.68 provides an equally effective method for preventing inadvertent criticality events in nuclear power plants, the NRC concludes that the final rule will not have any significant impact on the quality of the human environment.

Therefore, an environmental impact statement has not been prepared for this regulation. This discussion constitutes the environmental assessment for this rulemaking.

VII. Paperwork Reduction Act Statement This final rule does not contain a new or amended information collection requirement subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). Existing requirements were approved by the Office of Management and Budget, approval numbers 3150-0009 and 3150-0011.

VIII. Public Protection Notification If an information collection does not display a currently valid 0MB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

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IX. Regulatory Analysis The current structure of the current 10 GFR 70.24 is overly broad and places a burden on a licensee to identify those areas or operations at its facility where the requirements are unnecessary, and to request an exemption if the licensee has sufficient reason to be relieved from the requirements. This existing structure has resulted in a large number of exemption requests.

To relieve the burden on power reactor licensees of applying for, and the burden on the NRG of granting exemptions, this amendment permits power reactor facilities with nominal fuel enrichments no greater than 5-weight percent of U-235 to be excluded from the scope of 10 GFR 70.24, provided they meet specific requirements being added to 10 GFR Part 50. This amendment is a result of the experience gained in processing and evaluating a number of exemption requests from power reactor licensees and NRG's safety assessments in response to these requests which co'ncluded that the likelihood of criticality was negligible.

The only other viable option to this amendment is for the NRG to make no changes and allow the licensees to continue requesting exemptions. If no changes are made, the licensees will continue to incur the costs of submitting exemptions and NRG will incur the costs of reviewing them. Under this rule, an easing of the burden on licensees results from not having to request exemptions. Similarly, the NRG's burden will be reduced by avoiding the need to review and evaluate these exemption requests.

This rule is not a mandatory requirement,. but an easing of burden action which results in regulatory efficiency. Also, the rule does not impose any additional costs on existing licensees and has no negative impact on public health and safety, but will provide savings to Mure licensees, and may provide some reduction in burden to current licensees whose current 11

exemption includes conditions which are more restrictive than the requirements in Section 50.68. There will also be savings in resources to the NRC as well. Hence, the rule is shown to be cost beneficial.

The foregoing constitutes the regulatory analysis for this final rule.

X. Regulatory Flexibility Certification In accordance with the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b), the NRC hereby certifies that this rule, if adopted, will not have a significant economic impact on a substantial number of small entities. This rule affects only the licensees of nuclear power plants. These licensee companies that are dominant in their service areas, do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act, 5 U.S.C. 601, or the size standards adopted by the NRC (10 CFR 2.810).

XI. Backfrt Analysis The NRC has determined that this rule does not impose a backfrt as defined in 10 CFR 50.109(a)(1), since it provides an alternative to existing requirements on criticality monitoring.

Accordingly, the NRC has not prepared a backfrt analysis for this rule.

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XII.

• Small Business Regulatory Enforcement_ Fairness Act In accordance with the Small Business Regulatory Enforcement Fairness Act of 1996, 0

the NRC has determined that this action is not a "major rule and has verified this determination with the Office of Information and Regulatory Affairs, Office of Management and Budget.

XII I. List of Subjects 10 CFR Part 50 Antitrust, Classified information, Criminal penalties, Fire protection, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.

10 CFR Part 70 Criminal penalties, Hazardous materials transportation, Material control and accounting, Nuclear materials, Packaging and containers, Radiation_protection, Reporting.and recordkeeping requirements, Scientific equipment, Security measures, Special nuclear material.

For the reasons stated in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, the National Environmental Policy Act of 1969, as amended, and 5 U.S.C. 553, the NRC is adopting the following amendments to 10 CFR Parts 50 and 70:

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PART 50 DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES The authority citation for 10 CFR Part 50 continues to read as follows:

1. Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 Stat. 936, 937, 938,.

948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 2232,2233, 2236, 2239, 2282); secs.201, as amended, 202,206, 88 Stat. 1242, as amended 1244, 1246, (42 U.S.C. 5841, 5842, 5846).

Section 50.7 also issued under Pub. L. 95 - 601, sec. 10, 92 Stat. 2951, as amended by Pub. L. 102 - 486, sec. 2902, 106 Stat. 3123, (42 U.S.C. 5851). Section 50.10 also issued under secs. 101, 185, 68 Stat. 936, 955, as amended (42 U.S.C. 2131, 2235); sec. 102, Pub. L.

91 - 190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd), and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L. 91 - 190, 83 Stat. 853 (42 U.S.C. 4332).

Sections 50.34 and 50.54 also issued under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844).

Sections 50.58, 50.91, and 50.92 also issued under Pub. L. 97 - 415, 96 Stat. 2073 (42 U.S.C.

2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42 U.S.C. 2152). Sections 50.80 and 50.81 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234).

Appendix Falso issued under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).

2. Section 50.68 is added under the center heading "Issuance, Limitations, and Conditions of Licenses and Construction Permits" to read as follows:

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§ so.ea Criticality accident requirements.

(a) Each holder of a construction permit or operating license for a nuclear power reactor issued under this part or a combined license for a nuclear power reactor issued under Part 52 of this chapter, shall comply with either 10 CFR 70.24 of this chapter or the requirements in paragraph (b) of this section.

(b) Each licensee shall comply with the following requirements in lieu of maintaining a monitoring system capable of detecting a criticality as described in 10 CFR 70.24:

(1) Plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.

(2) The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.

(3) If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confiden~ level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.

(4) If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent 15

probability, 95 percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.

(5) The quantity of SNM, other than nuclear fuel stored onsite, is less than the quantity, necessary for a critical *mass.

(6) Radiation monito~ are provided in storage and associated handling areas when fuel is present to detect excessive radiation levels and to initiate appropriate safety actions.

(7) The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to five (5.0) percent by weight.

(8) The FSAR is amended no later than the next update which Section 50.71 (e) of this part requires, indicating that the licensee has chosen to comply with Section 50.68(b).

PART 70 DOMESTIC LICENSING OF SPECIALNUCLEAR MATERIAL The authority citation for 10 CFR Part 70 continues to read as follows:

1. Authority: Secs. 51, 53, 161, 182, 183, 68 Stat. 929, 930, 948, 953, 954, as amended, sec. 234, 83 Stat. 444, as amended, sec. 1701, 106 Stat. 2951, 2952, 2953 (42

.U.S.C. 2071, 2073, 2201, 2232, 2233, 2282, 2297f); secs. 201, as amended, 202, 204, 206, 88 Stat. 1242, as amended, 1244, 1245, 1246, (42 U.S.C. 5841, 5842, 5845, 5846).

Sections 70.1 (c) and 70.20a(b) also issued under secs. 135, 141, Pub. L. 425, 96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161). Section 70.7 also issued under Pub. L. 95 - 601, 16

sec. 10, 92 Stat. 2951 (42 U.S.C. 5851). Section 70.21(g) also issued under sec. 122, 68 Stat.

939 (42 U.S.C. 2152). Section 70.31 also issued under sec. 57d, Pub. L. 93 - 377, 88 Stat. 475 (42 U.S.C. 2077). Sections 70.36 and 70.44 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234).

Section 70.61 also issued under secs. 186, 187, 68 Stat. 955 (42 U.S.C. 2236, 2237).

Section 70.62 also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138).

2. In§ 70.24, paragraph (d) is revised to read as follows:

§ 70.24 Criticality accident requirements.

(d)(1) The requirements in paragraphs (a) through (c) of this section do not apply to a holder of a construction permit or operating license for a nuclear power reactor issued under Part 50 of this chapter or a combined license issued under Part 52 of this chapter, if the holder complies with the requirements of paragraph (b) of 10 CFR 50.68.

(2) An exemption from Section 70.24 held by a licensee who thereafter elects to comply with requirements of paragraph (b) of 10 CFR 50.68 does not exempt.thatlicensee from complying with any of the requirements in Section 50.68, but shall be ineffective so long as the licensee elects to comply with Section 50.68.

. ~

Dated at Rockville, Maryland this 28 day of Oct. , 1998.

William D. Trav rs Executive Dire or for Operations 17

[7590-01-P]

NUCLEAR REGULATORY COMMISSION DOCKETED USNRC 10 CFR Parts 50 and 70

'98 FEB 20 P2 :54 RIN: 3150-AF87 Criticality Accident Requirements; Withdrawal of Direct Final (jf~~:..: j/*_. *:-., :.:~:.:; y 1 ' ,.,.J,...*;-. :: -_;,--,.,cF Al""IJLJ0';,,..,.hi,~_

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Revocation of Regulatory Text AGENCY: Nuclear Regulatory Commission. DOCftET M1tBER PROPOSED RULE Pl 50 "'10 ,

( t,a. FR/,,38,;J.5)

( le>'2~~fo3,11)

ACTION: Direct final rule; withdrawal.

SUMMARY

The Nuclear Regulatory Commission is withdrawing a direct final rule that would have amended the Commission's regulations to provide light-water nuclear power reactor licensees with greater flexibility in meeting the requirement that licensees authorized to possess more than a small amount of special nuclear material (SNM) maintain a criticality monitoring system in each area where the material is handled, used, or stored. The NRC Is taking this action because it has received significant adverse comments in response to an identical proposed rule which was concurrently published in the Federal Register. Because the effective date for the direct final n.* :a has passed, the NRC is removing the regulatory text codified in ti ,at action.

J ~ ;/0 l'/'18 EFFECTIVE DATE: (Yf)OP puglicatign IA the Federal Reyis'9,j FOR FURTH'::R INFORMATION CONTACT: Stan Turel, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555. Telephone (301) 415-6234 (E-mail: spt@nrc.gov).

(Jµ},. ,.,,, a/t1t,/lJ'8' at tg3FI< '1 '/-03.,

2 SUPPLEMENTARY INFORMATION: On December 3, 1997 (62 FR 63825}, the Nuciear Regulatory Commission published in the Federal Register a direct finaJ rule amending its regulations to provide persons licensed to construct or operate light-water nuclear power reactors with the option of either meeting the criticality accident requirements of paragraph (a) of 10 CFR 70.24 in handling and storage areas for SNM, or electing to comply with requirements that would be incorporated into 10 CFR Part 50 at § 50.68. The direct final rule was to become effective on February 17, 1998. The NRC also concurrently published an identical proposed rule on December 3, 1997 (62 FR 63911 ). In these documents, the NRC indicated that If it received significant adverse comments in response to this action, the NRC would withdraw the direct final rule and would consider the comments received as in response to the proposed rule and address these comments in a subsequent final rule. Therefore, the Commission is withdrawing the December 3, 1997, direct final rule. The public comments received will be addressed in a subsequent final rule issued in either a notice of final rulemaking or in a notice of withdrawal of the proposed rule.

Because this notice of withdrawal is being published after the February 17, 1998, effective date for the direct final rule, the regulatory text presented in the December 3, 1997, direct final rule must be removed from the Code of Federal Regulations. Therefore, the provisions added to Part 50 at§ 50.68 are removed and the text of§ 70.24 (d) is being restored to the text of the paragraph that was in effect before the December 3, 1997, amendment to that paragraph.

3 List of Subjects 10 CFR Part 50 Antitrust, Classified information, Criminal pena~. Fire protection, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.

10 CFR Part 70 Criminal penalties, Hazardous materials transportation, Material control and accounting,

• Nuclear materials, Packaging and containers, Radiation protection, Reporting and recordkeeping requirements, Scientific equipment, Security measures, Special nuclear material.

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 adopting the following amendments to 10 CFR Parts 50 and 70.

PART 50 - DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES

1. The authority citation for Part 50 continues to read as follows:

AUTHORITY: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C.2132,2133,2134,2135,2201,2232,2233,2236,2239,2282);secs.201,as amended, 202, 206, 88 Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846).

Section 50.7 also issued under Pub. L. 95--601, sec. 10, 92 Stal 2951 (42 U.S.C.

5851). Section 50.10 also issued under secs. 101, 185, 68 Stat. 955 as amended (42 U.S.C. 2131, 2235), sec. 102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, and 50.54(dd), and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 50.55a and Appendix Q also issued under

4 sec. 102, Pub. L.91-190, 83 Stat 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58, 50.91, and 50.92 also Issued under Pub. L.97-415, 96 Stat. 2073 (42 U.S.C. 2239). Section 50.78 also Issued under sec. 122, 68 Stat. 939 (42 U.S.C. 2152). Sections 50.80- 50.81 also issued under sec. 184, .68 Stat. 954, as amended (42 U.S.C. 2234). Appendix Falso issued under sec. 187, 68 Stat. 955 (42 U.S.C 2237).

§ 80.68 (Removed]

2. Section 50.68 ls removed.

PART 70 DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL w

• 3. The authority citation for Part 70 continues to read as follows:

AUTHORITY: Secs. 51, 53, 161, 182, 183, 68 Stat. 929,930,948,953,954, as amended, sec. 234, 83 Stat. 444, as amended, (42 U.S.C. 2071, 2073, 2201, 2232, 2233, 2282, 2297f); secs. 201, as amended, 202,204, 206, 88 Stat. 1242, as amended, 1244, 1245, 1246 (42 U.S.C. 5841, 5842, 5845, 5846). Sec. 193, 104 Stat. 2835 as amended by Pub. L.

104-134, 110 Stat. 1321, 1321-349 (42 U.S.C. 2243).

Sections 70.1 (c) and 70.2Qa(b) also issued under secs. 135, 141, Pub. L.97-425, 96 Stat 2232, 2241 (42 U.S.C. 10155, 10161). Section 70.7 also issued under Pub. L.95-601, sec. 10, 92 Stat. 2951 (42 U.S.C. 5851). Section 70.21(9) also issued under sec. 122, 68 Stat.

939 (42 U.S.C. 2152). Section 70.31 also issued under sec. 57d, Pub. L.93-377, 88 Stat. 475 (42 U.S.C. 2077). Sections 70.36 and 70.44 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). Section 70.61 also issued under secs. 186, 187, 68 Stat. 955 (42 U.S.C. 2236, 2237). Section 70.62 also issued under sec. 108, 68 Stat. 939, as amended (42 u.s.c. 2138).

4. In § 70.24, paragraph (d) is revised to read as follows:

§ 70.24 Crtticallty accident requirements.

(d) Any licensee who believes that good cause exists why he should be granted an exemption in whole or in part from the requirements of this section may apply to the

5

_ Commission for such exemption. Such application shall specify his reason for the relief requested.

Dated at Rockville, Maryland, this 20th day of February, 1998.

For the Nuclear Regulatory Commission.

Jo~,&_

Secretary of the Commission .

DOCKETED USNRC January 13, 1998 "98 JAN 15 P4 :55 Secretary, U.S. Nuclear Regulatory Commission Attention: Rulemakings and Adjudications Staff DOCKET NlN9ER Washington, DC 20555-0001 PROPOSED RULE Pl SD,10

( <-2.F~ &.3t.a 6j

( t,'a.F"ll. fo3 et II)

Dear Sirs:

Proposed Rule Chance Issues The following are current rulemaking issues that may have an impact on the Trojan Nuclear Plant operations, procedures, and insurance requirements:

RlN 3150-AF87 (FR 62, No. 232,page 63825, dated December 3, 1997)

"Criticality Accident Requirements" The final rule is stated to become effective February 17, 1997, if no significant adverse comments are received.

Comment: A surface reading of this rule change implies that the proposed rule would be applicable to Trojan. By letter dated, February 16, 1993, PGE, however, had requested an exemption to the requirements of Part 70.24(a) and by_ letter, dated March 24, 1993 the NRC Staff responded that an exemption was not required because the requirements of Part 70.24(a) did not apply to the Trojan Plant Since the previous actions by the NRC Staff relate to the applicability of the current rule to the Trojan facility, and the rule change is forward looking, to reduce the level ofNRC Staff actions for plant specific exemptions to 10 CFR 70.24, the Trojan staff is of the opinion that the rule change is not intended to apply to plants similar to Trojan. It is recommended that the proposed*

rule be revised to clarify applicability for plants that have received NRC Staff actions (e.g.,

exemptions or other clarifying letters). Specifically, the final rule should have a provision that excludes from the scope of the rule any facility that has received NRC Staff action related to the application of 10 CFR 70.24(a).

• It should be noted that the criteria for determining that the Part 70.24(a) requirements did not apply to Trojan in the March 24, 1993 NRC letter are slightly different than the new FEB - 3 1998 Acknowledged by card ....... , "" . ' -

-'. .,., l requirements included in the proposed 10 CFR 50.68 that would form the basis for making Part 70.24(a) not applicable for shutdown and operating plants.

• The new criteria are not particularly difficult to implement (if we understand them correctly to not relate to cask movement evolutions), but they would require some procedure revisions and implementation of additional controls that are not currently required (e.g., items b. l, b.5, and b.6 of the proposed 50.68). The 'back.fit analysis' section of the proposed rule making does not reflect these addition costs. The Trojan facility is interested in minimizing cost for changes, particularly ones that have limited safety implications, since additional costs may impact the funds available for the decommissioning of the facility.

If there are any questions related to these comments, please contact Mr. H. R Pate at (503) 556-7480 or Mr. C. J. Stephenson at (503) 556-7465.

• (Retrieved from interactive rulemaking website -- ATB)

Co1T111enter:

Carl Stephenson Portland General Electric 71760 Columbia River Highway Rainier, OR 97048

Station Support Department PECO NUCLEAR DOCKETED L'SHRC PECO Energy Company 965 Chesterbrook Boulevard A Unit of PECO Energy Wayne, PA 19087-5691

• "98* JAN 14 P2 :53 January 7, 1998 Mr. John C. Hoyte Secretary of the Commission DOCKET PROPOSED Nll4BER RULE Pl So '1 7 o.

Attn: Rulemaklngs and Adjudications Staff ( <,Q F/U,JB~6)

U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

(<.,~ Fii. t,3'111)

Subject Comments Concerning Proposed and Direct Final Rules 10 CFR 50 and 70, "Crttlcallty Accident Requirements*

(62FR63911 and 62FR63825, dated December 3, 1997)

Dear Mr. Hoyle:

This letter is being submitted In response to the NRC's request for comments concerning Proposed and Direct Final Rules 10 CFR 50 and 70, "Criticality Accident Requirements,D which were published In the Federal Register (I.e., 62FR63911 and 62FR63825, dated December 3, 1997). This rulemaklng effort Involves changes to the NRC's regulations concerning criticality accident monitoring requirements for Special Nuclear Material (SNM). This rulemaking Is intended to.provide nuclear power reactor licensees with greater flexiblllty in meeting the requirement that licensees authorized to possess mOfB than a small quantity of SNM maintain a criticality monitoring system in each area where the material Is handled, used, or stored.

PECO Energy appreciates the opportunity to provide comments on the Proposed and Direct Final Rules. PECO Energy Is not opposed to promulgatlon of this Direct Final Rule; however, we believe that additional clarification with respect to the monitoring of unirradlated fuel in storage and associated handling areas is recommended. Specifically, unlrradiated fuel~ run require monitoring provided the fuel Is enclosed within an NRC-approved (I.e., 10CFR71, "Packaging and Transportation of Radioactive Materiar) shipping package. The entire approved shipping package typically consists of two (2) rectangular boxes comprised of an outer wooden container and an inner metal container, which houses the fuel. There is only cushioning material between the two (2) containers. The containers are designed In accordance with a Certtflcate of Compliance (COC) for radioactive materials packages Issued by the NRC for shipment of unlrradlated fuel assemblies. The COC recognizes that both outer and Inner containers comprise the approved package. The Inner metal box Is the container that ensures that a geometrically safe configuration of fuel is maintained during transport, handling, storage, and accident conditions, and that the Introduction of any moderating agents to the fuel is precluded due to its leak-tight construction. Inadvertent criticality is prevented due to the construction of the container and the storage configuration of the fuel in the container. Therefore, we recommend clarification with regard to the monitoring requirements for when the fuel has been removed from the out~r wooden container, but Is still within the metal container and being moved or transferred within the confines of the site boundary.

Acknowled 8d by FEB ~ 3 1998 g <:aJd **"*Hlll***::>orr ~r:r:c~

January 7, 1998 Page2 If you have any questions, please do not hesitate to contact us.

Very truly yours, G. A. Hunger, Jr.

Director - Licensing

0)

Fermi 2 DOCKETED ~f~L_,

Detroit 6400 North D1x1e Hwy USNRC Edison Newport, M1ch1gan 48166 (313) 586-5300 ~ Nuclear Generation "98 JAN -9 P4 :29 January 2, 1998 NRC-98-0012 DOCKETNlNBERa ,..

PROPOSED RULE r~ :J o J- 7O Secretary ( (p -:2. Ft<G.3'&~5)

U. S. Nuclear Regulatory Commission ( re, :J. FR. (p 3 9 It)

Washington D. C. 20555-0001 Attention: Rulemaking and Adjudications Staff

References:

1) Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43

2) NRC Letter~dated October 31, 1997 "Exemption From Criticality Accident Requirements In 10 CFR 70.24(a) - Grand GulfNuclear Station, Unit 1 (TAC NO. M96177)"

Subject:

Detroit Edison Comments on the Proposed and Direct Final Rulemaking on Criticality Accident Requirements, 10CFR Parts 50.68 and 70.24 (62 FR 63825 and 63911)

On December 3, 1997, the Nuclear Regulatory Commission (NRC) issued a proposed and direct final rule with opportunity to comment on Criticality Accident Requirements (62 FR 63825 and 63911). The purpose of this letter is to submit Detroit Edison's comments on the above rules.

The enclosure to this letter provides Detroit Edison's comments on the above subject rules. Detroit Edison is concerned that the proposed changes will not provide sufficient flexibility in meeting the regulations relating to criticality monitoring and will require Detroit Edison to request an exemption from the rules unless the comments are satisfactorily resolved and/or incorporated in the final rule prior to its proposed effective date of February 17, 1998.

JAN 2 9 1998 h;knowledged by c a r d ~

USNRC NRC-98-0012 Page2 If you should have any questions concerning Detroit Edison's comments please contact Harl 0. Arora, Principal Licensing Engineer, at (313 or 734) 586-4213.

Sincerely,

//~

Norman K. Peterson Director, Nuclear Licensing Enclosure cc: K. Cozens (NEI)

D. J. Modeen (NEI)

Enclosure to NRC-98-0012 Page 1 Comments on Final Rulemaking on Criticality Accident Requirements, 10CFR 50.68 and 70.24 10 CFR 50.68(b)(3)

The requirement for Keff<0.98 with optimum moderation of fresh fuel of maximum permissible U-235 enrichment loaded in the new fuel storage racks filled with low-density hydrogenous fluid cannot be met at some Boiling Water Reactors (BWRs).

General Electric (GE) dealt with this issue over 20 years ago, and concluded that there is an extremely remote possibility for inadvertently establishing critical conditions in the new fuel storage racks, or in a dry spent fuel pool loaded with new fuel.

An analysis by GE indicated that it would require the introduction of a low equivalent water-density material to completely occupy the space in and around an array of fuel assemblies in storage for the occurrence of a criticality. Both 10x25 and 20x25 bundle arrays were analyzed, with and without gadolinia, to simulate reactivity conditions from initial core loads to the most reactive design basis reload fuel (as of 1976). In all cases, the optimum moderation occurred when the equivalent water density was approximately equal to 0.2 gram/cc. In the worst case, a range of equivalent water densities from 0.05 to 0.45 grams/cc was undesirable in conforming to the 0.98 Keff design basis limit.

In the interest of assuring safety margins in the areas where fuel is handled, additional controls that further reduce the probability of a criticality occurrence were recommended by GE to their customers in Service Information Letter (SIL) -152 "Criticality Margins for Storage ofNew Fuel," dated March 31, 1976. In summary, the SIL recommends actions for keeping the new fuel storage vault dry (drains open, no fire protection fogging nozzles in the area etc.).

Detroit Edison believes that criticality in the new fuel storage racks is not a credible event provided utilities followed the guidance given in SIL 152 and the criteria in 10 CFR 50.68(b)(3) should be revised to include exemption from the requirements if administrative controls preclude optimum moderation conditions.

Enclosure to NRC-98-0012 Page 2 10 CFR 50.68(b)(6)

The NRC needs to define "Fuel Handling," and "Storage and Associated Handling Areas." This section requires that General Design Criteria (GDC) 63 be met. However, GDC 63 only addresses monitoring of the fuel storage and associated handling areas in terms of being in reactor refueling areas, and does not address the case when the fuel is unloaded at another location. This needs to be clarified whether this is only a requirement during fuel assembly movement or if it applies to movement of inner metal containers without the outer container.

The proposed changes to 10 CFR 50.68 do not address the recent issue that the GE inner RA3 metal container by itself is not considered to be an approved shipping container per 10 CFR 71, and therefore, the handling of the inner metal container without the outer wooden overpack falls under the 10 CFR 70.24 requirements. The proposed (1 0CFR 50.68 (b) (6)) rule does not clearly address the concern whether an approved shipping container (per 10 CFR 71) is required to prevent a criticality event. It is Detroit Edison's understanding that the inner container provides sufficient criticality protection. This agrees with the NRC statement as stated in an NRC grant of exemption (Reference 2) for Grand Gulfs 10 CFR 70.24 exemption request. In this grant of exemption, the NRC stated, "It is the inner metal container that ensures that a geometrically safe configuration of the fuel is maintained during transport, handling, storage, and accident conditions, and that the introduction of any moderating agents to the fuel is precluded due to its leak-tight construction."

We suggest revising 10 CFR 50.68 (b)(6) to read, " ... associated handling areas when fuel assemblies are removed from the approved metal containers per 10 CFR 71 to detect. .. "

DOCtCET Nll&R PROPOSED RULE PR 5o .1- .zq

( f# ~ F=~ <, 38 :J 5') DOCKETED From:

( <P ::l Flt. (, J '1 II)

"Dewhirst, Linda R.* <lrdewhi@nppd.com>

us~mc To : "'CAG@nrc.gov** <CAG@nrc.gov>

Date: 1/8/98 6:53pm '98 JAN -9 P4 :25

Subject:

Additional Comments on Final Rule 10 CFR 50.68 Ms. Gallagher: OFF!(_.~:  :: ,-.

RULU:, , ,

Recognizing the below comments are after the requested time but I ~~);'~A to share them anyway and ask for feedback if possible. I"ve heard that other utilities have similar issues. (I'm having trouble with my web browser recently so I thought I would take the email route). Thank you.

Comments on 10CFRS0.68, 10CFR70.24 Direct Final Rulemaking:

50.68(b) is unclear. What is the definition of transportation? Does this mean as soon as the truck which is hauling the numerous bundles of new fuel enters the restricted (protected) area (fuel is in an approved transportation container at this point)? The regulation does not say.

It would be ridiculous for us to perform a determination on this truck

• under the most adverse moderation conditions feasible by unborated water* if the bundles are still in their transportation container. Has the GE container truly been analyzed under the most adverse conditions feasible up to the point the bundle is unloaded from the box? What does handling *at any one time* mean? Does this mean that I can't unload one box from the truck on one elevation while operators are inspecting a bundle in the inspection stand on the refuel floor because I don't have a "determination" covering the most adverse moderation conditions? How is the *most adverse moderation conditions feasible by unborated water defined.* What is considered an acceptable "determination?*

50.58 (bl (2) and (3) are silent about storing the new fuel on the refueling floor rather than in the new fuel storage vault (we do not use ours and when ITS goes into effect, it's prohibited). How will we be affected?

50.68 (b) (5) is very vague. Under the right conditions (i.e., in a laboratory environment) very small quantities of SNM could be made critical. Laboratory conditions are not applicable in our case but yet we are limited to "less than the quantity necessary for a critical mass.* Why didn't the NRC add the criteria from Reg Guide 10.3 which is very specific in its definition and is more applicable to power reactors (which are the intended audience for this regulation)?

50.68 (b) (7)--why are we limiting enrichment? Why not keep it to Keff being less than our limit?

The regulation is silent regarding licensees who already have an approved exemption request to 10CFR70.24 from the NRC. In addition, several utilities received an exemption before the seven criteria were published in IN 97-77 (CNS is not among them however, nor do we have an exemption at this time)

We are in the process of developing our exemption request; however if 50.68 is promulgated as planned, then is this necessary (providing we meet the requirements of the rule, see the issues above).

Happy New Year!

JAN 2 9 1998 owledg cara **-----AA..,-.,.....,

Linda R. Dewhirst Licensing Engineer Cooper Nuclear Station Tel: 402.825.5009 Fax: 402.825.5827 email: lrdewhiinppd.com

IJfl llO.C.KEIE Northem -1111~ mmDCompany Monticello Nuclear Generating Plant 2807 West HW\:.7.9 .\.~ ~

DOCKET NUMBER* ...

PROPOSED RULE rn !JDJ 10 Montice118 Miq)Mo~3Jl.W$"l le, A,,~~, Jr8

( <pp_ Flt.'43 8~5) OFFl*:..:1.- 1F SF,::,: ; HY AUL::'.,;.** *'J' - ~. 'I;::;

January 2, 1998

( (, ~ FR "1 3 , II) ADJUDib*- _, .*AFF Secretary U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 Attn: Rulemaking and Adjudications Staff The following comments are respectively submitted in response to the proposed changes to Criticality Accident Requirements, 10 CFR 50.68 and 70.24, published in Federal Register Vol ume 62, Number 232, Page 63825, December 3, 1997.

The phrase "as required by GDC 63" of proposed 10 CFR 50.68 (b) (6) should be removed for the following reasons. First, some plants were licensed before the General Design Criteria were promulgated and their licensing bases address the GDC on a case-by-case basis; the phrase in question infers that the General Design Criteria as stated in 10 CFR Part 50 Appendix A are part of every licensees' design basis. Second, the phrase does not add any substance since proposed 50. 68 (b) (6) simply restates the relevant portion of GDC 63; omitting the reference would be consistent with proposed 50.68 (b) (1) through (5) which implement GDC 62 without specific reference to that GDC. Third, a person unfamiliar with 10 CFR 50 Appendix A would not recognize the reference to GDC 63 as stated .

Proposed 10 CFR 50.68 (b) (7), which places a five (5.0) weight percent limit on U-235 enrichment, should be eliminated and the phrase "maximum permissible U-235 enrichment" in proposed 50.68 (b) (2), (3), and (4) should be replaced by the phrase "maximum fuel assembly reactivity" for the following reasons. First, the discussion in the Federal Register announcement does not indicate that the enrichment limitation is the basis for a safety analysis; it is simply a statement of current practice. Second, the safety issue is fuel assembly reactivity of which enrichment is only one parameter; burnable poison , material selection , and geometry are major factors affecting reactivity that could compensate for higher enrichments. Third, by modifying 50.68 (b) (2), (3), and (4) as proposed, the reactivity limitation objective of fuel storage racks can be achieved without placing a limitation on fuel enrichment.

AcknOWledged by card*--~~~-~- ..__998

• We appreciate the opportunity to comment on this proposed rule change.

1J lf D~

Marcus H. Voth Project Manager - Licensing cc: T J Kim, NRR-PM, NRC Kris Sanda J E Silberg

DOCKETED USNRC N U C L E A R E N E RGY I N ~ T ILJAlf -5 pJ :Q5 OFFICE_ ~ ;F sr=c R[-- -,__:.-~Qavld J. Modeen AUL 1::~ .' **,i"-1*

l.l\ '

- " -. . 'DIRECTOR, ENGINEERING H\JU NUCLEAR GENERATION DIVISION ADJ UD!C. , ;:~TAFF January 2, 1998 DOCKET NlllBER Secretary U.S. Nuclear Regulatory Commission PROPOSED RULE Pl 5 0 .J 7Q

( (p :J F~ t, 39' :J.5) .

Washington, DC 20555-0001

( (,:JF~ f,39/lj

SUBJECT:

Comments on the Criticality Accident Requirements Proposed and Direct Final Rulemaking (62 Fed. Reg. 63825 and 63911)

Enclosed are the Nuclear Energy Institute (NEI} 1 comments on the Criticality Accident Requirements proposed and direct final rulemaking (62 Fed. Reg. 63825 and 63911). The new §50.68 and the revised §70.24 are scheduled to become effective February 17, 1998, unless significant adverse comments are received by the NRC. Our review has identified several issues that represent significant adverse comments. NEI requests that the NRC not proceed with the direct final rule, but instead follow an expedited schedule to resolve comments on the proposed rule.

The rulemaking objective to eliminate the need for a significant number of exemption requests pursuant to §70.24 is appropriate and will be achieved if the rule is amended to address industry's comments. By contrast, the §50.68(b)(3) requirement, as written, will require a significant number of licensees to submit exemption requests. This paragraph establishes a new requirement for fresh fuel storage racks that might inadvertently be filled with low-density hydrogenous fiuid, such as fog or sprays. Since 1976, BWR licensees have managed this concern with administrative controls consistent with those described in GE SIL 152, Criticality Margins for Storage of New Fuel. Paragraph (b)(3) should be revised to allow licensees an option to minimize the likelihood and impact of low-density hydrogenous fiuid using administrative controls. An in-depth discussion of concern with §50.68(b)(3) is contained in Enclosure 1.

Acknowfedged by card ...~~N ? 9 1998 1 NEI is the organization responsible for establishing unified nuclear industry policy on matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEI's members include all utilities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect/engineering firms, fuel fabrication facilities, materials licensees, and other organizations and individuals involved in the nuclear energy industry.

l 776 I STREET , NW SUITE 400 WASHINGTON, DC 20006-3708 PHONE 202 739 8000 FAX 202 785.4019

Secretary, U.S. NRC January 2, 1998 Page2 Neither the rule no;r its statement of consideration explicitly address the status of existing §70.24(d) exemptions. Nothing in the rule should invalidate exemptions previously granted by the NRC. The rule needs clarification on the status of existing exemptions otherwise licensees may re-submit exemptions previously approved by the NRC. The NRC staff should amend the rule to affirm the continuing validity of existing exemptions. This will assure that neither the NRC nor industry needlessly waste resources.

The rule should clarify the relationship of Part 71 shipping container handling requirements to §50.68 and §70.24. One could interpret these regulations to mean that when an inner metal shipping container is removed from its outer wood container that the provisions of Part 71 are not satisfied and that handling of the inner metal container alone will require management per the requirements of

§50.68 or §70.24. It is the inne,: metal container that provides criticality protection.

The outer wood box is not necessary to prevent criticality. An NRC letter to the Grand Gulf Nuclear Station dated October 31, 1997 states, "It is the inner metal container that ensures that a geometrically safe configuration of fuel is maintained during transport, handling, storage and accident conditions ... " Sections 50.68 and 70.24 should be amended to clearly state that there is no need for criticality monitoring when handling the inner metal container without its wood over pack.

This will eliminate the likelihood of licensees submitting exemption requests to continue use of their current fuel handling practices.

Enclosure 2 provides additional comments necessary to clarify the rule.

The proposed rule should be implemented only after these comments are addressed.

Licensees are likely to submit numerous exemption requests to the NRC if the rule remains as written.

• If you have questions concerning our comments, please contact Kurt Cozens at (202) 739-8085 or koc@nei.org.

Sincerely, / / . / / / /J c;:sY4 ~

David J. Modeen KOC/edb Enclosures c: Stan Tu.rel, NRC/NRR S. Singh Bajwa, NRC/NRR

ENCLOSURE 2 ADDITIONAL COMMENTS ON THE DIRECT FINAL RULE COMMENT# PARAGRAPH COMMENT

1. 10CFR50.68(b)(1) The paragraph reads, "Plant procedures may not permit handling and transporlation at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.*

a) Recommend replacing the phrase "may not permit" with the phrase "shall prohibit the" to express this as a clear requirement b) The terms "handling and transportation* and "safely subcrltlcal" should be expllcltly defined to avoid misinterpretation.

c) Revise the paragraph to clarify that the determination is to be made by the license, such as In an engineering calculation.

cf) In Paragraphs 10CFR50.68(b)(2) and 10CFR50.68(b){4), the moderator is identified as "pure water" rather than "unborated water,* as described In 10 CFR 50.68(b)(1). If the moderators we~ intended to be the same, then the paragraphs should be revised to use the same words. Otherwise, some explanation of the difference between "pure water" and "unborated water" might be necessary to avoid Mure misunderstandings.

e) Paragraph 10CFR50.68(b)(3) discusses "optimum moderation" by a "low-density hydrogenous fluid." The phrases "most adverse moderation* and "optimum moderation" seem to express opposite retatlonshlps, but are used to describe the same physical phenomenon. 'Mlen an assumption of low-density hydrogenous fluid is required for the optimum moderation for new fuel storage, clarification Is necessary to understand the basis for using unborated water to determine the most adverse moderation for handling and transportation.

2. 10CFR50.68(b)(2) The proposed paragraph reads, in part, 7he estimated ratio of neutron production to neutron absorption and leakage (k-effectlve) of the fresh fuel ....

Since all neutrons (that are produced) subsequently either leak or are absorbed, the paragraph should be clarified to specify its applicability to an instant In time. Alternately, the paragraph may be revised to eliminate the words "ratio of neutron production to neutron absorption and leakage,* since "k-effective* Is a sufficiently understood term to permit Its use without the need to define it.

3. 10CFR50.68(b)(2) These paragraphs address fresh fuel storage racks, but at least one" and licensee has committed not to use such storage racks In order to avoid 10CFR50.68(b)(3) criticality accident concerns. For simplicity, these paragraphs should be revised to be applicable unless the license institutes administrative controls to prohibit the use of fresh fuel storage racks.

4. 10CFR50.68(b)(5) The paragraph reads, 7he quantity of SNM, other than nuclear fuel stored on site, is less than the quantity necessary for a critical mass.*

There could be a situation where widely scattered sources on sight would add up to a crltlcal mass. If these widely scattered SNM sources are part of the fuel or handled like fuel, they should not be considered part of the total for the same reason that fuel ls not Plant procedures and controls for SNM are adequate to control accident criticality. The paragraph shouJd be revised to reflect this situation.

COMMENT# PARAGRAPH COMMENT

5. 10CFR50.68(b)(6) The paragraph reads, "Radiation monitors, as required by GOG 63, are provided in storage and associated handling areas when fuel is present to detect excessive radiation levels and to initiate appropriate safety actJons.

• a) To be pree1se, GDC 63 requires that appropriate systems be provided to detect excessive radiation levels and to initiate appropriate safety actionS'. Logically, radiation monitors would be a necessary part of such systems, but GDC 63 does not require the radiation monitors to initiate safety actions. This paragraph should be clarified.

b) Some plants were not licensed to the 1971 General Design criteria, but were licensed under other criteria. The paragraph should be r8Vlsed to reflect the license conditions. Revise the paragraph to ellminate reference to GDC 63 and describe the underlying monitoring requirements or to require "Radiation monitors, as required by GOG 63 or other analogous 1/censee criteria, ... "

c) The requirement that "Radiation monitors ... are provided in storage and associated handling areas .. ." is Inappropriate. Fuel storage areas for both new and used fuel are not normally occupied volumes. As such, not all fuel storage volumes (vaults or pools) have radiation monitoring Inside of them. In some cases, monitoring 1s located outside of the storage volume to monitor conditions within the storage volume. The paragraph should be changed such that "In" Is replaced with "in the vicinity of.*

cf) Use of the wording* ... initiate appropriate safety actions* Is inappropriate. At some facilities, these detectors are not formally classified as safety related. The paragraph should be revised t9 replace "initiate appropr:urte safety actions* with "initiate appropriate warning."

6. 10CFR50.68(b)(7) The paragraph reads, 'The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to no greater than five (5.0) percent by we,ghl" This requirement is unnecessary and precludes the development of advanced fuel designs. Any changes In enrichment above 5.0 percent by weight would be supported by an updated cntJcality analysis for both dry and spent fuel racks to ensure the appropriate margins to criticality are maintained. Placing a limit on enrichment provides no direct safety benefit and should not be included. The explicit numerical criteria should be eliminated from the rule.

ENCLOSURE 1 SIGNIFICANT ADVERSE COMMENT ON 10 CFR 50.68{b)(3)

Current BWR fuel storage racks may not comply with the Kea<0.98 requirement of

§50.68(b)(3). These licensees would them need to submit an exemption request, unless the requirement is revised to permit administrative controls such as those identified in GE SIL-152, Criticality Margins for Storage of New Fuel (Attachment A). Presently licensees are managing the §50.68(b)(3) concern with administrative controls.

Licensees and GE concluded in 1976 that an extremely low probability exists for inadvertently establishing critical conditions with fresh fuel in the new fuel storage racks or in a dry spent fuel pool. SIL-152 states that criticality could not be, achieved without the introduction of a low equivalent water density material to completely occupy the space around an array of fuel assemblies. GE fuel bundle arrays were analyzed, with and without gadolinia, to simulate reactivity conditions from initial core loads to the most reactive design basis reload fuel (as of 1976). In all cases, the optimum moderation occurred when the equivalent water density was approximately equal to 0.2 gram/cc. In the worst case loading con.figuration, equivalent water densities from 0.05 to 0.45 gram/cc, the BWR fuel storage arrangement may not comply with the proposed Kea<0.98 requirement. Licensees believe that administrative controls recommended in the SIL are appropriate to manage the Keff concern. The NRC staff was informed of the SIL recommendations at the time of its issuance and did not disagree. Licensee have been using the SIL guidance since 1976.

Criterion 50.68(b)(3) should be revised to include an exemption from the requirements if administrative controls preclude optimum moderation conditions.

Attachment A

-U. *O!Q:rmatior letts1 March 31, 1976 SIL Ho. 152 File Tab A Category 1 CRITICALITY MARGINS FOR STORAGE OF NEW RJEL Using optimum moderatpr conditions, calculations indicate that there fs an extremely remote possibility for inadvertently establishing critical conditions in the new fuel storage racks, or in a dry spent fuel pool loaded with new fuel.* Potential sources for an optimum moderator are nre extinguisher foam, water mist, steam or other hydrogenous materials.

This Service Infonnation Letter (SIL) recorrmends precautionary measures tu BWR operators to further reduce the already very low probability of sw::.h an event occurring.

DISCUSSION An.analysis by General Electric indicated that it would require the introduction of a low eiiuivalent water density material to completely_

~cupy .the space in ~nd around an.arrav of n,el-assemblies i~ stora~e r the occurrence of .a crf tita 1i ty. Both 10 x 25 and 20 x* 2~ bund e arrays were analyzed. with and without gadolinia, to simulate reactivity conditions from initfal-core loads* to the most reactive design basis reload fuel. In all cases the optimum 100der.at f~n occurred when the equivalent water 9ensi.ty was

• 0.2 gram/cc. In* the worst ca~e. a range of equivalent water densities from.0.05 to 0.45 irams/cc was undesirable in confonnfng to the design basis Keff lim1ts. This concern Has been judged by General Electric not to be a reportable deficiency and the judgment has been supported by the NRC.

Resu1t~. of BWR _site s.urveys ha'{e indicated the presence.of fire hose stations and sprinklers on the refueling flonr at a significant nUll'lber of plants. ~lso, a substantial nwnber of hoses in these stations are provided with adjustable nozzles *that are variable rrom a solid stream to a coarse spray. In the interest of assuring safety margins on the refueling floor, additional controls that further reduce the probability of a criticali.ty occurrence should be implemented.

RECOMMENDED ACTI01l General Electrfc reccmmends that the procedural controls listed below be considered at the earliest convenient opportunity to reduce the nnote probability.for inadveftently establishing critical conditions for new fuel stora~: * *** ~ * * .. *

• NUCLE..\R eNERGV DIVISION ,. OPEAAmfG PLANT SERVICES

• SAN JOSE, CALtFOilMIA SS125 NO WARRANTY 0. RePRESINTATI.Otll &XMESSiD ,OR IWU!D S MAOE WITH RISP&CT TO THE ACCURACY. COMPLETl!:NESll OR tlnl'uutes OP THIS INFOR*

u.-ncm. OEN!AAI.. l!LECTFHC COM,.AN'r' A.."'SUM!:S NO REff'OHStBIUTY POR 1.lA*

BILITY OR DAMAGE WHICH MAY RESULT FROM THE USi OF nus INFOR-...a.TION.

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• r 6 EMERA l L:,; :I fl 1:.

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March 31. 1976 SIL No. 152 Attachment PROCEDURAL RECDrflEtmATI0NS FOR NORMAL FUEL HANOUf1G OPEAAllI0NS

1. No more than one fuel bundle should be suspendec above the fuel storage array and this at a height no greater than 24 inches to limit penetration displacement if the bundle was dro?ped.

2. Fuel handling in the fuel storage area should be limited tc one fuel assembly or the weight equivalent per crane. An exception to this requirement is a properly designed fuel shipping container or an overload test weight. The shipping container or overloa.d test weight should at not;~~ be suspended above the fuel storage array.

• I 3: A fuel array.of up to three fuel bundles outside of a nonnal storage area or nonna1 *shipping container should be maintained with an edge-to-edge spacing of 12 inches or more from all other fuel.

• I * *

4. A fuel array of four or more fuel bundles outside of the nonnal fuel storage areas or properly designed fuel shipping container should be prohibite*.

5. The new fuel vault should always be dry.

6. The spent fuel pool .should be maintained 1n a-flooded condition if new fuel is in storage during construction activities or: when construction debris are pre~ent. Flooding .should provf d~ ~t least

• enough water to cover the bundles. If the spent fuel pool is not flooded when new fuel is in storage, the fuel should be covered by a solid, firepr.oof material. to prevent possible inundation by low density fire extinguisher foam or water mist.

7. New fuel should'not be stored such that a fuel bundle could remain flooded without water existing between bundles.

8. Fuel movement 1n the new fuel vault should not be permitted if an abnonnal condition of vault flooding occurs.

9. Fuel should not be placed in aisles or moved through aisles adjacent to and at the same level of the storage racks.

10. Defective fuel should always be stored in defective fuel storage containers and placed in the defective fuel storage rack or control rod storage rack.

11. If fuel 1s stored in temporary storage racks below the fuel pool.

work table, the work table .should not ba used to handle fuel; conversely, if the'work table is used to handle fuel, fuel storag~

below the work table should be prohibited. .

March 26. 1976 SIL No. 152 Attachment

12. No more than two fuel bundles should be allowed 1n or around a fuel prep machine at any ~ime. This fuel should be separated from the main body of stored_fluel by at least 12 inches.

13. Fuel should not be stored outside of designated storage*cells.

14. New fuel should not be stgred in the nlN fuel vault when there are construction activities on the refueling f100r or construr.tfon debris in the vicinity of the new fu~l vault unless a solid, fireproof cover is placed over the fault which would preclude criticality due to inundation by low density water such as water fog or spray from a fire hose.

March 31, 1976 SIL No. 152

1. The new fuel storage vault should always be dry. For exanPt)le, it should be iDJPOssible to block the drain. or in_any:way produce the equivalent water densities in the ranges noted above.

2. The spent fuel pool should be flooded or covered with a fireproof cover if new fuel is in storage when construction actfvitfes or construction debri's are present. Flooding should provide at least enough water to cove?' the bundles. In taking these steps, the plant owner should be careful to assur~ that the fuel pool cooli,ng system is either inoperative or properly vented prior to startup to preclude an air cleating event through the spargers. The dispersion of many small bubbles is a potential source of low equivalent water density.

3. Fuel should not be stored in the nl!W fuel vault when there are construction act1v1t1es on the refueling floor or construction debris in tne vicinity of the new fuel vault unless a so1id

~ '!r is placed over the vault. This solid cover would 11e-(p to prevent the introduction of low density water such as a fog or spray should the operation of fire hoses become necessary on the refueling floor.

4. The attachment to this SIL @ntitled, "Procedural Reconnendations for Normal Fuel Handling Operations" should be reviewed by

~ plant personnel to assure a complete understanding of all current procedural controls relevant to fuel handling operations.

It should be noted that for reasons of emphasis items 5, 6 and-14i~ the ~ttachment are identical to the procedural control recommendations l. 2 and 3 listed above.

For additional information and assistance, consult your local General Electric service representative.

Prepared by: C. J, Paone/L. A. Gonzalez Approved by: ~i.t.aytt:;/Manag2r o:c. t. L- Issued by:

anager Product Service Service Connun1cat1ons Product

Reference:

A71

• Plant Recomnendations Jll - Fuel and Reloads

JAf'l. 2. 1gg8 2: 45PM SOUTHERN t'lUCLEHR 1 205 J 3'::lc b rn:::.

Lawis Somner Sautltam Nacr..r Vice President Dpel'ltln, Compqy, lni;.

Hatch Project Support 40 Inverness Pamwy oocr;Ei ED PDst Office Box 1295 USNRC Birmingham. Alabama 35201 Tel 205.992.7279 Fax zos 992.0341 December 31, 1997 Docket Nos. s0 ..321 50-348 50-424 HL-5546 50-366 S0-364 50-425 LCV-1145 Mr. John C. Hoyle Secretary of the Commission DOCKET Nt.1tBER U.S. Nuclear Regulatocy Commission Washington, DC 20SSS-0001 PROPOSED RULE flR 5 0 .J 7 0

( r,~ F~ t,39~5)

( (, ~ F~ ltP 3 'III)

ATTENTION: Ruleroaking and Adjudications Staff Comments on Direct Final Rule "Criticality Accident Requirements

(62 Federal Reaister 63825 dated December 3, 1997)

Dear Mr. Hoyle:

On December 3, 1997, the Nuclear Regulatory Commission (NRC) published in the Federal Register concUirently as a proposed rule (62 FR 63911) and as a direct final rule (62 FR 63825) with opportunity to comment, changes to the regulations on criticality accident requirements contained in 10 CFR S0.68 and 10 CFR 70.24. In accordance with

• the request for comments, Southern Nuclear Operating Company is in total agreement with the Nuclear Energy Institute comments which are to be provided to the NRC regarding this issue.

Should you have any questions, pl~ advise.

Respectfully submitted, H. L. Sumner, Jr.

HLS/TMM (distribution - see next page)

Acknowledged by card ..JAN 2 9 1998

I JAN. 2.1998 2: 46PM SOUTHERN NUCLEAR ( 205 ) '3"32 6108 U. S. Nuclear Regulatory Commission Page2 cc: Southern Nt\dem Q.peratini Company Mr. D. N. Morey, Vice President, Plant Farley Mr. C. K. McCoy, Vice President, Plant Vogtle Mr. 1. B. Beasley, General Manager - Plant Vogtle Mr. R. D. Hill, General Manager* Plant Farley Mr. P.H. Wells, General Manager -Plant Hatch u, s, Nuolear Rc'"Iatgcy Comm,ission,$ashinpm, pc Mr. J. I. Zimmerman, Licensing Project Manager

• Farley Mr. N. B. Le, Licensing Project Manager* Hatch Mr. D. H. Jaffe, Senior Project Manager* Vogtle u, s, Nuclear R&&ulatory Commission. Reinon II Mr. L. A Reyes, Regional Administrator Mr. T. M. Ross, Senior Resident Inspector - Farley Mr. B. L. Holbrook, Senior Resident Inspector - Hatch Mr. J. Zeiler, Senior Resident Inspector- Vogtle HL-5S46 LCV-1145

CP&L 0

Carolina Power & Light Company PO Box 1551

.411 Fayetteville Street Mall Raleigh NC 27602 CP&L Letter: PE&RAS-97-101 December 24, 1997 DOCKET NUMBER PROPOSED RULE Pft 5 O+70

( (p:J.F~e,~s~s)

( "~F,e ~3t:tll)

Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Attn: Rulemakings and Adjucations Staff

Subject:

Comments on NRC Proposed and Direct Final Rules on 10CFRS0.68 and 10CFR70.24 Criticality Accident Requirements (62 FR 63825 and 62 FR 63911)

Dear Sir or Madam:

Attached are the comments of Carolina Power & Light Company (CP&L) on the NRC Proposed and Direct Final Rules on 10CFR50.68 and 10CFR70.24 Criticality Accident Requirements. In general, CP&L supports this change as an efficient and effective improvement in the regulatory process.

Please contact me at (919) 546-6901 if you have questions.

Sincerely, D.B. Alexander, Manager Perfonnance Evaluation & Regulatory Affairs HAS Attachment AcknOWledged 0y can1----

DEC 3 1 \997

Page 2 CP&L Letter PE&RAS-97-101 December 24, 1997 Comments on NRC Proposed and Direct Final Rules on 10CFRS0.68 and 10CFR70.24 Criticality Accident Requirements (62 FR 63825 and 62 FR 63911) cc: Mr. L.J. Callan, Executive Director for Operations Mr. S.J. Collins, Director, USNRC Office of Nuclear Reactor Regulation Mr. L.A. Reyes, Regional Administrator, Region II Mr. J.B. Brady, USNRC Resident Inspector - HNP, Unit 1 Mr. B.B. Desai, USNRC Resident Inspector- HBRSEP, Unit 2 Mr. V.L. Rooney, USNRC Project Manager- HNP, Unit 1 Ms. B.L. Mozafari, USNRC Project Manager- HBRSEP, Unit 2 Mr. C.A. Patterson, USNRC Resident Inspector - BSEP, Units 1 and 2 Mr. D.C. Trimble, USNRC Project Manager - BSEP, Units 1 and 2 Chairman J.A. Sanford - North Carolina Utilities Commission USNRC Document Control Desk

Page 3 CP&L Letter PE&RAS-97-101 December 24, 1997 Attachment Comments on NRC Proposed and Direct Final Rules on 10CFRS0.68 and 10CFR70.24 Criticality Accident Requirements (62 FR 63825 and 62 FR 63911)

1. The proposed paragraph 10CFR50.68(b)(l) reads:

"Plant procedures may not permit handling and transportation at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water. "

a) In order to express this as a clear requirement, CP&L suggests replacing the phrase "may not permit" with the phrase "shall prohibit the."

b) CP&L suggests that the paragraph be revised to clarify that the determination is to be made by the license, in a engineering calculation for example, rather than by the NRC in a Safety Evaluation.

c) In subsequent paragraphs 10CFR50.68(b)(2), 10CFR50.68(b)(3) and 10CFR50.68(b)(4), subcriticality is expressed as a maximum limit (either 0.95 or 0.98 or 1.0) on the estimated k-effective at a 95 percent probability and a 95 percent confidence level. Since the requirement is "to be safely subcritical," is 1.0 the correct maximum limit on k-effective? Or, does the absence of specific criteria imply the application of a different standard? CP&L suggests that more specific criteria be added to paragraph 10CFR50.68(b)(l).

d) In subsequent paragraphs 10CFR50.68(b)(2) and 10CFR50.68(b)(4), the moderator is identified as "pure water" rather than "unborated water." If the moderators under consideration were intended to be the same, then CP&L suggests that these paragraphs be clarified to use the same words. Otherwise, some further explanation of the difference between "pure water" and "unborated water" might be necessary to avoid future misunderstandings.

e) Paragraph 10CFR50.68(b)(3) discusses "optimum moderation" by a "low-density hydrogenous fluid." The phrases "most adverse moderation" and "optimum moderation" seem to express opposite relationships but are used to describe the same physical phenomenon. CP&L suggests some clarification is necessary. CP&L also suggests that some clarification is necessary to help understand why it is appropriate to use unborated water to determine the most adverse moderation for handling and transportation when an assumption of a low-density hydrogenous fluid is required for the optimum moderation for new fuel storage.

2. The proposed paragraph 10CFR50.68(b)(2) reads, in part:

"The estimated ratio ofneutron production to neutron absorption and leakage (k-effective) of the fresh fuel .... "

Since all neutrons (that are produced) subsequently either leak or are absorbed, CP&L suggests that the paragraph be clarified to specify its applicability to an instant in time.

Alternately, CP&L suggests that paragraph be revised to eliminate the words "ratio of neutron production to neutron absorption and leakage," since "k-effective" is a sufficiently understood term to permit its use without the need to define it.

Page 4 CP&L Letter PE&RAS-97-101 December 24, 1997 Attachment Comments on NRC Proposed and Direct Final Rules on 10CFRS0.68 and 10CFR70.24 Criticality Accident Requirements (62 FR 63825 and 62 FR 63911)

3. Paragraphs IOCFR50.68(b)(2) and 10CFR50.68(b)(3) address fresh fuel storage racks, but CP&L understands that at least one licensee has committed not to use such storage racks in order to avoid criticality accident concerns. For simplicity, CP&L suggests that these paragraphs be revised to be applicable unless the license institutes administrative controls to prohibit the use of fresh fuel storage racks.

4. The proposed paragraph 10CFR50.68(b)(6) reads:

"Radiation monitors, as required by GDC 63, are provided in storage and associated handling areas when fuel is present to detect excessive radiation levels and to initiate appropriate safety actions. "

To be precise, GDC 63 requires that appropriate systems be provided to detect excessive radiation levels and to initiate appropriate safety actions. Logically, radiation monitors would be a necessary part of such systems, but GDC 63 does not require the radiation monitors to initiate safety actions. CP&L suggests that this paragraph be clarified.

5. The proposed paragraph 10CFR50.68(b)(7) reads:

"The maximum nominal U-235 enrichment ofthefreshfuel assemblies is limited to no greater thanfive (5.0) percent by weight. "

CP&L understands that at least one U.S. reactor is currently pursuing a license to operate with test assemblies containing mixed-oxide fuel. Until either more operating experience or more analysis is available for MOX fuel, CP&L suggests that this paragraph be revised to limit the fissionable material to U-235.

DOCKETED USNRC "97 IIC 29 P3 :19 CP&L Letter: PE&RAS-97-101 December 24, 1997 DOCKET NlltBER -~

PRoPoseo RULE Pl

{ ~:2.F~ (.18'~ 5')

so 1 o

( ft, :J. F/f! ts,~ q 11)

Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Attn: Rulemakings and Adjucations Staff

Subject:

Comments on NRC Proposed and Direct Final Rules on 10CFR50.68 and 10CFR70.24 Criticality Accident Requirements (62 FR 63825 and 62 FR 63911)

Dear Sir or Madam:

Attached are the comments of Carolina Power & Light Company (CP&L) on the NRC Proposed and Direct Final Rules on 10CFR50.68 and 10CFR70.24 Criticality Accident Requirements. In general, CP&L supports this change as an efficient and effective improvement in the regulatory process.

Please contact me at (919) 546-6901 if you have questions.

Sincerely, Original signed P.A. Opsal for D.B. Alexander

[received on interactiv~ rulemaking website on 12/24/97 - ATB]

D.B. Alexander, Manager Performance Evaluation & Regulatory Affairs HAS Attachment Acknowfedged

Page CP&L Letter PE&RAS-?7-101 2 December 24, 1997 Comments on NRC Proposed and Direct Final Rules on 10CFR50.68 and 10CFR70.24 Criticality Accident Requirements (62 FR 63825 and 62 FR 63911) cc: Mr. L.J. Callan, Executive Director for Operations Mr. S.J. Collins, Director, USNRC Office of Nuclear Reactor Regulation Mr. L.A. Reyes, Regional Administrator, Region II Mr. J.B. Brady, USNRC Resident Inspector- HNP, Unit I Mr. B.B. Desai, USNRC Resident Inspector - HBRSEP, Unit 2 Mr. V.L. Rooney, USNRC Project Manager - HNP, Unit 1 Ms. B.L. Moz.afari, USNRC Project Manager - HBRSEP, Unit 2 Mr. C.A. Patterson, USNRC Resident Inspector- BSEP, Units I and 2 Mr. D.C. Trimble, USNRC Project Manager- BSEP, Units 1 and 2 Chairman J.A. Sanford - North Carolina Utilities Commission USNRC Document Control Desk

Page 3 CP&L Letter PE&RAS-97-101 December 24, 1997 Attachment Comments on NRC Proposed and Direct Final Rules on 10CFRS0.68 and 10CFR70.24 Criticality Accident Requirements (62 FR 63825 and 62 FR 63911)

1. The proposed paragraph 10CFR50.68(b)(1) reads:

"Plant procedures may not permit handling and transportation at any one time ofmore fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water. ,,

a) In order to express this as a clear requirement, CP&L suggests replacing the phrase "may not permit" with the phrase "shall prohibit the."

b) CP&L suggests that the paragraph be revised to clarify that the determination is to be made by the license, in a engineering calculation for example, rather than by the NRC in a Safety Evaluation.

c) In subsequent paragraphs 10CFR50.68(b)(2), 10CFR50.68(b)(3) and 10CFR50.68(b)(4), subcriticality is expressed as a maximum limit (either 0.95 or 0.98 or 1.0) on the estimated k-effective at a 95 percent probability and a 95 percent confidence level. Since the requirement is to be safely subcritical," is 1.0 the correct maximum limit on k-effective? Or, does the absence of specific criteria imply the application of a different standard? CP&L suggests that more specific criteria be added to paragraph 10CFR50.68(b)(l).

d) In subsequent paragraphs 10CFR50.68(b)(2) and 10CFR50.68(b)(4), the moderator is identified as "pure water" rather than "unborated water." If the moderators under consideration were intended to be the same, then CP&L suggests that these paragraphs be clarified to use the same words. Otherwise, some further explanation of the difference between "pure water" and unborated water" might be necessary to avoid future misunderstandings.

e) Paragraph 10CFR50.68(b)(3) discusses "optimum moderation" by a "low-density hydrogenous fluid." The phrases "most adverse moderation" and "optimum moderation" seem to express opposite relationships but are used to describe the same physical phenomenon. CP&L suggests some clarification is necessary. CP&L also suggests that some clarification is necessary to help understand why it is appropriate to use unborated water to determine the most adverse moderation for handling and transportation when an assumption of a low-density hydrogenous fluid is required for the optimum moderation for new fuel storage.

2. The proposed paragraph 10CFR50.68(b)(2) reads, in part:

"The estimated ratio ofneutron production to neutron absorption and leakage (k-effective) of the fresh fuel .... "

Since all neutrons (that are produced) subsequently either leak or are absorbed, CP&L suggests that the paragraph be clarified to specify its applicability to an instant in time.

Alternately, CP&L suggests that paragraph be revised to eliminate the words "ratio of neutron production to neutron absorption and leakage," since "k-effective" is a sufficiently understood term to permit its use without the need to define it.

Page 4 CP&L Letter PE&RAS-97-101 December 24, 1997 Attachment Comments on NRC Proposed and Direct Final Rules on 10CFRS0.68 and 10CFR70.24 Criticality Accident Requirements (62 FR 63825 and 62 FR 63911)

3. Paragraphs IOCFR50.68(b)(2) and IOCFR50.68(b)(3) address fresh fuel storage racks, but CP&L understands that at least one licensee has committed not to use such storage racks in order to avoid criticality accident concerns. For simplicity, CP&L suggests that these paragraphs be revised to be applicable unless the license institutes administrative controls to prohibit the use of fresh fuel storage racks.

4. The proposed paragraph IOCFR50.68(b)(6) reads:

"Radiation monitors, as required by GDC 63, are provided in storage and associated handling areas when fuel is present to detect excessive radiation levels and to initiate appropriate safety actions. "

To be precise, GDC 63 requires that appropriate systems be provided to detect excessive radiation levels and to initiate appropriate safety actions. Logically, radiation monitors would be a necessary part of such systems, but GDC 63 does not require the radiation monitors to initiate safety actions. CP&L suggests that this paragraph be clarified.

5. The proposed paragraph IOCFR50.68(b)(7) reads:

"The maximum nominal U-235 enrichment of the.fresh fuel assemblies is limited to no greater than jive (5. OJ percent by weight. "

CP&L understands that at least one U.S. reactor is currently pursuing a license to operate with test assemblies containing mixed-oxide fuel. Until either more operating experience or more analysis is available for MOX fuel, CP&L suggests that this paragraph be revised to limit the fissionable material to U-235.

December 29. 1997 NOTE TO: Emile Juli an Chief. Docketing and Services Branch FROM: Carol Gallagher

/, ~.:.,,.if~._,,,

RES. ORA l b--_,J_

SUBJECT:

DOCKETING OF COMMENT ON DIRECT FINAL/PROPOSED RULEMAKING Attached for docketing is a coll1llent letter related to the Direct Final/Proposed Rulemaking on Criticality Accident Requirements. This letter was received on our interactive rulemaking website on December 24. 1997. The commenter's name and address are D.B. Alexander. Carolina Power &Light Co ..

Raleigh, NC 27601. Please send a copy of the docketed comment to Stan Turel (mail stop T9-F-31) for his records.

Attachment:

As stated cc w/o attachment:

S. Turel

( omm,1'1\Halth Edi,on < i11p,11, I -100 <>pu~ Pla<:l Do,, ner, <,ro, c: II <,o.:;' .:;_.:;- 1 OOCK E"fEO

([)

US~RC "97 OEC 29 P2 :53 ComEd December 22, 1997 DOCKETNlA18ER ., .

U.S. Nuclear Regulatory Commission ATTN: Rulemaking and Adjudications Staff PROPOSED RULE Pfl 5 O J-10 Washington, DC 20555-0001 ( ~ ~ F~ t,JB~s)

( ~ ~F~ t-3,1!)

Subject:

Comments on Direct Final Rulemaking Criticality Accident Requirements, 10CFR Parts 50.68 and 70.24

Reference:

Federal Register (FR) Vol. 62, No. 232 dated December 3, 1997.

This letter provides the Commonwealth Edison Company (ComEd)'s comments on the subject Nuclear Regulatory Commission (NRC) proposed rulemaking. The comment period for this Direct Final Rule expires on January 2, 1988.

ComEd's comments are provided in the Attachment.

Please provide any questions you may to this office.

/J ri

/~~

Thomas I. Kovach/'-

Vice President Nuclear Regulatory Services Attachment cc: G. Dick, Generic Issues Project Manager - NRR A. B. Beach, Regional Administrator - RIII Office of Nuclear Safety - IDNS DEC 3 1 1997 Acknowledged by card' "".......................a,., *.

\ l ni<:om Compan~

ATTACHMENT Comments on Proposed Direct Final Rule Criticality Accident Requirements 10CFR70.24 - Criticality Accident Requirements Proposed Change (d) The requirements in paragraph (a) through (c) of this section do not apply to holders ofa construction permit or operating license for a nuclear power reactor issued pursuant to part 50 of this chapter, or combined licenses issued under part 52 of this chapter, if the holders comply with the requirements ofparagraph (b) of JO CFR 50.68 of this chapter.

Comments:

The current version of 10CFR70.24(d) contains provisions for applying for exemptions should "good cause" exist. ComEd is concerned that the proposed change impacts the ability to apply for such exemptions. This is of particular concern because ComEd has pending 10CFR70.24 exemptions with the Commission. The proposed rule should not prohibit licensees from applying for such exemptions under the guidelines of IOCFR70.14. In addition, the new rule should contain provisions to note that any existing approved exemptions remain valid.

10CFR50.68(b)- Criticality Accident Requirements Proposed Rule (b) Each licensee shall comply with the Jo/lawing requirements in lieu of maintaining a monitoring system capable of detecting a criticality as described in JO CFR 70.24:

(J) Plant procedures may not permit handling and transportation at any one time of more fuel assemblies than have been determined to be safely subcritica/ under the most adverse moderation conditions feasible by unborated water.

Comments:

ComEd has no comments on this section.

1

ATTACHMENT Comments on Proposed Direct Final Rule Criticality Accident Requirements (2) The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum permissible U-235 enrichment and.flooded with pure water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level.

Comments:

ComEd has no comments on this section.

(3) If optimum moderation offresh Juel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum permissible U-235 enrichment and filled with /ow-density hydrogenous fluid, the k-effective co"esponding to this optimum moderation must not exceed 0.98, at a-95 percent probability, 95 percent confidence level.

Comments:

For the ComEd Boiling Water Reactors (BWRs), optimum moderation calculations are not performed for the fresh fuel storage racks. It is our understanding that this is the case for many BWRs. In accordance with vendor recommendations, compensatory measures have been established to preclude an optimum moderation condition in the fresh fuel storage racks. The rule should include a provision that exempts this requirements if adequate controls have been established to preclude an optimum moderation condition.

(4) If no credit/or soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum permissible U-235 enrichment must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, ifflooded with pure water. If credit is taken for soluble boron, the k-effective of the spent Juel storage racks loaded with fuel of the maximum permissible U-235 enrichment must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, ifflooded with borated water, and the k-effective must remain below 1. 0 (subcritica/), at a 95 percent probability, 95 percent confidence level, ifflooded with pure water.

Comments:

ComEd has no comments on this section.

2

ATTACHMENT Comments on Proposed Direct Final Rule Criticality Accident Requirements (5) The quantity of SNM, other than nuclear fuel stored on site, is less than the quantity necessary for a critical mass.

Comments:

ComEd has no comments on this section.

(6) Radiation monitors, as required by GDC 63, are provided in storage and associated handling areas when fuel is present to detect excessive radiation levels and to initiate appropriate safety actions.

Comments:

ComEd operates several facilities that were licensed prior to formal adoption of the General Design Criteria in I 0CFR50, Appendix A. Although the intent of GDC 63 is being maintained at these facilities, a literal read of this requirement may conclude that reactors licensed prior to the adoption of the GDCs could not meet this requirement. The Rule should eliminate the reference to GDC 63 and describe the underlying monitoring requirements.

ComEd has concerns over the wording "Radiation monitors ...are provided in storage and associated handling areas ...". Fuel storage areas for both new and used fuel are not normally occupied volumes. As such, not all ComEd fuel storage volumes (vaults or pools) have radiation monitoring inside of them. In some cases, monitoring is located outside of the storage volume to monitor conditions within the storage volume. Therefore, the Rule should be changed such that "in" is replaced with "in the vicinity of'.

In addition, ComEd has concerns over the wording " ... .initiate appropriate safety actions". At some ComEd facilities, these detectors are not formally classified as safety related. The Rule should be changed such that "initiate appropriate safety actions" is replaced with "initiate appropriate warning."

3

ATTACHMENT Comments on Proposed Direct Final Rule Criticality Accident Requirements (7) The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to no greater than.five (5.0) percent by weight.

Comments:

This requirement is unnecessary and precludes the development of advanced fuel designs. Any changes in enrichment above 5.0 percent by weight would be supported by an updated criticality analysis for both dry and spent fuel racks to ensure the appropriate margins to criticality are maintained. Placing a limit on enrichment provides no direct safety benefit and should not be included.

4

DOCKETED USNRC

[7590-01-P]

"97 OEC -9 AlO :35 NUCLEAR REGULATORY COMMISSION 10 CFR Parts 50 and 70 RIN: 3150-AF87 Criticality Accident Requirements AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

SUMMARY

The Nuclear Regulatory Commission (NRC) is amending its regulations to provide light-water nuclear power reactor licensees with greater flexibility in meeting the requirement that licensees authorized to possess more than a small amount of special nuclear material (SNM) maintain a criticality monitoring system in each area where the material is handled, used, or stored. This action is taken as a result of the experience gained in processing and evaluating a number of exemption requests from power reactor licensees and NRC's safety assessments in response to these requests that concluded that the likelihood of criticality was negligible.

~ ~ J'l'lf' DATES: Comments on the proposed rule must be received 0,1 or before ~O ea~& eftel"'

publicafioo in tt::ia ~aaaral Regi&&ar).

ADDRESSES: Mail comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 , Attention: Rulemaking and Adjudication Staff. Hand deliver comments to 11555 Rockville Pike, Maryland, between 7:45 am and 4:15 pm on Federal workdays.

Copies of any comments received may be examined at the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC.

For information on submitting comments electronically, see the discussion under Electronic Access in the Supplementary Information section.

FOR FURTHER INFORMATION CONTACT: Stan Turel, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Wash;ngton, DC 20555-0001 , telephone (301) 415-6234, e-mail spt@nrc.gov.

SUPPLEMENTARY INFORMATION:

For additional information see the Direct Final Rule published in the rules section of this Federal Register.

Procedural Background Because NRC considers this action noncontroversial and routine, we are publishing this proposed rule concurrently as a direct final rule. The direct final rule will become effective on

, a . ~ '71 19'1~

('iti davs aflor pt:1blieation in ti 1e Fede, al l'teglster). However, if the NRC receives significant

}~u *""f" R, l'l"f ~

adverse comments on the direct final rule by (30 days after FJWblieatieA iA ~ho i;:aaoral Regis&or)1 then the NRC will publish a document that withdraws the direct final rule. If the direct final rule is withdrawn, the NRC will address in a Final Rule the comments received in response to the proposed revisions in a subsequent final rule. Absent significant modifications to the proposed revisions requiring republication, the NRC will not initiate a second comment period for this action in the event the direct final rule is withdrawn.

Electronic Access You may also provide comments via the NRC's interactive rulemaking web site through the NRC home page (http://www.nrc.gov}. This site provides the availability to upload comments as files (any format}, if your web browser supports that function. For information about the interactive rulemaking site, contact Ms. Carol Gallagher, (301} 415-6215; e-mail CAG@nrc.gov.

List of Subjects 10 CFR Part 50 Antitrust, Classified information, Criminal penalties, Fire protection, Intergovernmental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.

10 CFR Part 70 Criminal penalties, Hazardous materials transportation, Material control and accounting, Nuclear materials, Packaging and containers, Radiation protection, Reporting and recordkeeping requirements, Scientific equipment, Security measures, Special nuclear material.

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, the National Environmental Policy Act of 1969, as amended, and 5 U.S.C. 553, the NRC is considering adopting the following amendments to 10 CFR Parts 50 and 70.

PART SO-DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES The authority citation for 10 CFR Part 50 continues to read as follows:

1. Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 Stat. 936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 2132, 2133,2134,2135,2201,2232,2233,2236,2239,2282);secs.201,asamended,202,206,88 Stat. 1242, as amended 1244, 1246, (42 U.S.C. 5841, 5842, 5846).

Section 50.7 also issued under Pub. L.95-601, sec. 10, 92 Stat. 2951, as amended by Pub. L. 102 - 486, sec. 2902, 106 Stat. 3123, (42 U.S.C. 585.1). Sections 50.10 also issued under secs. 101, 185, 68 Stat. 936, 955, as amended (42 U.S.C. 2131, 2235); sec. 102, Pub. L.

91 - 190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd), and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L. 91 -190, 83 Stat. 853 (42 U.S.C. 4332).

Sections 50.34 and 50.54 also issued under sec. 204, 88 Stat. 1245 (42 U.S.C. 5844).

Sections 50.58, 50.91, and 50.92 also issued under Pub. L. 97 - 415, 96 Stat. 2073 (42 U.S.C.

2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42 U.S.C. 2152). Sections 50.80 50.81 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234). Appendix Falso issued under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).

2. Section 50.68 is added under the center heading "Issuance, Limitations, and Conditions of Licenses and Construction Permits" to read as follows:

§ so.as Criticality accident requirements.

(a) Each holder of a construction permit or operating license for a nuclear power reactor issued under this part, or a combined license for a nuclear power reactor issued under part 52 of this chapter shall comply with either 10 CFR 70.24 of this chapter or requirements in paragraph (b).

(b) Each licensee shall comply with the following requirements in lieu of maintaining a monitoring system capable of detecting a criticality as described in 10 CFR 70.24:

(1) Plant procedures may not permit handling and transportation at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.

(2) The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum permissible U-235 enrichment and flooded with pure water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level.

(3) If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum permissible U-235 enrichment and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level.

(4) If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum permissible U-235 enrichment must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with pure water. If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum permissible U-235 enrichment must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percer.t confidence level, if flooded with pure water.

(5) The quantity of SNM, other than nuclear fuel stored on site, is less than the quantity necessary for a critical mass.

(6) Radiation monitors, as required by GDC 63, are provided in storage and associated handling areas when fuel is present to detect excessive radiation levels and to initiate appropriate safety actions.

(7) The maximum nominal U-235 enrichment of the fresh fuel assemblies is limited to no greater than five (5.0) percent by weight.

PART 70-DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL The authority citation for 10 CFR Part 70 continues to read as follows:

1. Authority: Secs. 51, 53, 161, 182, 183, 68 Stat. 929,930,948, 953, 954, as amended, sec. 234, 83 Stat. 444, as amended, sec. 1701, 106 ~tat. 2951, 2952, 2953 (42 U.S.C.2071,2073,2201,2232,2233,2282,22970;secs.201,asamended,202,204,206, 88 Stat. 1242, as amended, 1244, 1245, 1246, (42 U.S.C. 5841, 5842, 5845, 5846).

Sections 70.1(c) and 70.20a(b) also issued under secs. 135, 141, Pub. L. 97 -425, 96 Stat. 2232, 2241 (42 U.S.C. 10155, 10161). Section 70.7 also issued under Pub. L. 95- 601, sec. 10, 92 Stat. 2951 (42 U.S.C. 5851). Section 70.21(g) also issued under sec. 122. 68 Stat. 939 (* .C. 2152). Section 70.31 also I s * * ~

• sec. 57d, Pub. L.93-377. 88 Stat. 475 (42 U.S.C. 2077). Sections 70.36 and 70.44 also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234).

Section 70.61 also issued under secs. 186, 187, 68 Stat. 955 (42 U.S.C. 2236, 2237).

Section 70.62 also issued under sec. 108, 68 Stat. 939, as amended (42 U.S.C. 2138).

2. In§ 70.24, paragraph (d) is revised to read as follows:

§ zo.24 CriticalitY accident regu;rements.

(d) The requirements in paragraph (a) through (c) of this section do not apply to holders of a construction permit or operating license for a nuclear power reactor issued pursuant to part 50 of this chapter, or combined licenses issued under part 52 of this chapter, if the holders comply with the requirements of paragraph (b) of 10 CFR 50.68 of this chapter.

Dated at Rockville, Maryland this Lt.."-day of Nd(/ , 1997.

For the Nuclear Regulatory Commission.

L Joseph Callan, Execu

• e Director for Operations.