ML23152A144

PR-050, 070 - 62FR63825 - Criticality Accident Requirements

ML23152A144
Person / Time
Issue date:	12/03/1997
From:	Callan L NRC/EDO
To:
References
PR-050, PR-070, 62FR63825
Download: ML23152A144 (1)

Text

ADAMS Template: SECY-067

DOCUMENT DATE: 12/03/1997

TITLE: PR-050, 070 - 62FR63825 - CRITICALITY ACCIDENT REQUIREMENTS

CASE

REFERENCE:

PR-050, 070

62FR63825

KEYWORD: RULEMAKING COMMENTS

Document Sensitivity: Non-sensitive - SUNSI Review Complete STATUS OF RULEMAICING

PROPOSEP RULE: PR-050, 070 OPEN ITEM (Y/N) N

RULE NAME: CRITICALITY ACCIDENT REQUIREMENTS

PROPOSED RULE FED REG CITE: 62FR63825

PROPOSED RULE PUBLICATION DATE: 12/03/97 NUMBER OF COMMENTS: 9

ORIGINAL DATE POR COMMENTS: 01/02/98 EXTENSION DATE: I I

FINAL RULE FED. REG. CITE: 63PR63127 FINAL RULE PUBLICATION DATE: 11/12/98

NOTES ON: WOULD ALLOW LICENSEES MORE FLEX. IN MTNG. REQUI. OP MAINT'NG CRITI STATUS : CALITY MONITORING SYST. FOR SNM. /S/'D BY EDO. SEE PR AT 62FR63911 OP RULE:

• DFR WITHDRAWN - 63FR9402, PUB. 2/25/98.

HISTORY OF THE RULE

PART AFFECTED: PR-050, 070

RULE TITLE: CRITICALITY ACCIDENT REQUIREMENTS

PROPOSED RULE PROPOSED RULE DATE PROPOSED RULE SECY PAPER: SRM DATE: I I SIGNED BY SECRETARY: I I

FINAL RULE FINAL RULE DATE FINAL RULE SECY PAPER: SRM DATE: I I SIGNED BY SECRETARY: 10/28/98

STAFF CONTACTS ON THE RULE

CONTACTl: STAN TUREL MAIL STOP: T-9F31 PHONE: 415-6234

CONTACT2: MAIL STOP: PHONE:

DOCKET NO. PR-050, 070 (62FR63825)

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 - DIRECT FINAL RULE WITH OPPORTUNITY TO COMMENT

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 l 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 OOCKET NUMBER

~ROPOSED RULE PR 5 0.1-10 ooct,E TEO

( (p 2 FR t, 3 i'1 i)' [7590-01-P] USNRC

'98 NOV 13 AB :20 NUCLEAR REGULATORY COMMISSION

10 CFR Parts 50 and 70

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.

~ !Lf, !998 EFFECTIVE DATE: The final rule is effective on... (30 de!I) ! efter publieetio11 i11 the Fede1 al

-Register~.

pµg_ ~ 1,f12jq8' df. C,:3FR tp 3 I :17 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 1 O CFR Part 50. The requirements in Section 50.68 are

generally the requirements that the NRC has used to grant specific exemptions from the

requlrer,:ients of 1 O 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 1 O CFR 70.24 do not appiy to holders of a construction permit or

operating license for a nuclear power reactor issued under 1 O CFR Part 50, or combined

licenses issued under 1 O CFR Part 52, if the holders comply with the requirements of 1 O CFR

50.68(b).

2

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 1 0 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.

3 111. 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 Cf R 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 10 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.

4 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 that 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."

5 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 Mure 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.

6 Comment 6. Use of "pure water" and "unborated wate'r" 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 storag~ racks are not

used."

Comment 8. The meaning of "transportation" 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 powe~ 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

50.6~. 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).

A 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 permit or a combined operating license) the option of complying

with Section 70.24(a) through (c), or com'plying with the 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 effectivl:3 method

tor 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 tor this rulemaking.

VI I. 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.

VII I. 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.

10 IX. Regulatory Analysis

The current structure of the current 10 CFR 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

NRC 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

CFR 70.24, provided they meet specific requirements being added to 10 CFR 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 NRC's safety assessments in response

to these requests which concluded that the likelihood of criticality was negligible.

The only other viable option to this amendment is for the NRC 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 NRC 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 NRC'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 ard safety, but will provide savings to future

licensees, and may provide some reduction in burden to current licensees whose current 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 backfit 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 backfit analysis for this rule.

12 XII. Small Business Regulatory Enforcement Fairness Act

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

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.

XIII. 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 amen-dments to 1 o C'FR Parts 50 and 70:

13 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, 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. 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

-S0.56 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 Pennits* to read as follows:

14

§ so.ea Criticality accident reguirements.

(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 1 0 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 confidence 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 confldence_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 monitors 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 SPECIAL NUCLEAR MATERIAL

The authority citation for 1 O 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); seCQ. 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. 97 - 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 that.licensee 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 REGULA TORY COMMISSION DOCKETED USNRC 10 CFR Parts 50 and 70

'98 FEB 20 P 2 :54 RIN: 3150-AFB?

Criticality Accident Requirements* ' Pf Withdrawal of Direct Final<j1/4/llCEcf,-~,r el~.!} VI,-.:- ' ii 8 ::\\.;! \\' (.::,,.;, I \\;*Ju 1L., /j~':RY ADJUD1CATIQ,, :: :3TAFF Revocation of Regulatory Text

AGENCY: Nuclear Regulatory Commission. ~~

PROPOSED RULE Pl 50 d-10

( t,a.FRi,38eJS)

ACTION: Direct final rule; withdrawal. ( ~':J.F~to3,11}

SUMMARY

The Nuclear Regulatory Commission is withdrawing a direct finar 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 materiel! (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 tl*,at

action.

a~ e:JS;,118 EFFECTIVE DATE: (wpon pulii,lisatigA IR tRe Federal Regi&IBtj

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).

{) µ),. In g_/:it5/C/fl o:t <i,3FI<. C/ '/-02-,

2

SUPPLEMENTARY INFORMATION: On December 3, 1997 (62 FR 63825), the Nuclear

Regulatory Commission published in the Federal Register a direct final 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 1 O 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.

I:

3

List of Subjects

10 CFR Part 50

Antitrust, Classified infonnatlon, 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, and 5 U.S.C

553, the NRC is adopting the following amendments to 10 CFR Parts 50 and 70.

PART 60 - DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES

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

AUTHORllY: 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 (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).

§ 50.68 [Removed]

2. Section 50.68 is removed.

PART 70 - DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL

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 70.20a(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(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 al~o 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 Critlcallty 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.

Far the Nuclear Regulatory Commission.

Jo~Le__

Secretary of the Commission.

DOC KE TUSNRC ED

January 13, 1998 '98 JAN 15 P4 :55

Trojan Nuclear Plan t O FFI C f., \\;-.. ~; * *. ;~ I *-. *f D k 50 344 RU I l f,,..,*.I', oc et - ADJU D' : /."'1*<*.. *,*.. -F License NPF-1

Secretary, U.S. Nuclear Regulatory Commission Attention: Rulemakings and Adjudications Staff Washington, DC 20555-000 1

Dear Sirs:

Proposed Rule Change Issues

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

RIN 31 50-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, ifno 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

cknowtedged by card............. F!B - 3 19.,,.,,,,s.,

U.S. OOLEAR REGULATORY CO lEMAKI GS & ADJU tCATIO

• OFFICEOFTHF OF THEC D I

Postmark Oat ~ ~, C~ 4,£(,,/....,. ~ 1/15"/t;e II I Rep 3

"° r. tribut: -rU/t.d1 7> D4 l'\\IC,s_

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.1, b.5, and b.6 of the proposed 50.68). The 'backfit 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)

Commenter:

Carl Stephenson Portland General Electric 71760 Columbia River Highway Rainier, OR 97048 January 14. 1998

NOTE TO: Emile Juli an Chief. Docketing and Services Branch FROM : Carol Gallagher RES. DRA ~f>>i

SUBJECT:

DOCKETING OF COMMENT ON DIRECT FINAL RULE

Attached for docketing is a comment letter related to the Direct Final

- Rule on Criticality Accident Requirements. This letter was received via the rulemaking webs i te on January 13. 1998. The commenter's name is Carl Stephenson. Portland General Electric. 71760 Columbia River Highway. Rainier.

OR 97048. 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 Station Support Department

- DOCKETED PECO NUCLEAR USN RC 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. Hoyle 5o.J1o Secretary of the Commission -*--""""!'--

Attn: Rulemakings and Adjudications Staff ( (,~ FR<P.38~5)

U.S. Nuclear Regulatory Commission (v'J. F/l. IP3°111)

Washington, DC 20555-0001

Subject:

Comments Concerning Proposed and Direct Final Rules 1 O CFR 50 and 70, "Criticality 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," which were published in the Federal Register (i.e., 62FR63911 and 62FR63825, dated December 3, 1997). This rulemaking 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 flexibility in meeting the requirement that licensees authorized to possess more than a small quantity of SNM maintain a criticality monitoring system in each area where the material is handled, used, or stored.

PECO Energy appreciat e s the opportunity to provide comments on the Proposed and Direct Final Rules. PECO Energy is not opposed to promulgation of this Direct Final Rule; however, we believe that additional clarification with respect to the monitoring of unirradiated fuel in storage and associated handling areas is recommended. Specifically, unirradiated fuel~ nQ1 require monitoring provided the fuel is enclosed within an NRG-approved (i.e., 10CFR71, "Packaging and Transportation of Radioactive Material") 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 Certificate of Compliance (COC) for radioactive materials packages issued by the NRC for shipment of unirradiated 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 outer wooden container, but is still within the metal container and being moved or transferred within the confines of the site boundary.

Acknowledaed by card.......... F'EB - 3 1998 ----

U.S. NUCLEAR REGULATORY COM ISSION RULEMAKINGS & ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION Ooc11T1ent Statistk:s

Postmark. Date I q /q&"

Copies Reoeived __ / ______ _

Add'I Copies ReproduceQ --==:.-r-= ~

Special Distnbution,_.-r._:,~...Jo!J,1,...-J;...-.~

r;? ""-=s _______ _

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 DOC KE TED Detroit 6400 North Dixie Hwy U S RC Newport, Michigan 48166,~, Nuclear Edison (313) 586-5300, r Generation "98 JAN -9 P 4 :29

January 2, 1998 NRC-98-0012 DOCKET PROPOSED RULE 5 °,;- 7 O Secretary ( (p :l Fl<G,3&- :1 5)

U.S. Nuclear Regulatory Commission ( '1 :J. Pll ~3'111)

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) - Gran d GulfNuclear Station, Unit 1 (TAC NO. M96177)"

Subject:

Detroit Edison Comments on the Proposed and Direct Final Rulemaking on Criticality Accident Requirements, 1 0CFR 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.

AcknoWledged by card.... JAN 2 9 998 ____ _

U.S. NUCLEAR REGULATORY COMMISSION RULEMAKINGS&AOJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION Oocllnant Statistics Postmmk Oete '/,;) ' I c; '{?'

CoplesRecalwai ___, __ _

Add~~ R&produced _ 3.____ __ _

Special '/<IDS I D1s1ributk>n ~ u rel, --PDR1 USNRC NRC-98-0012 Page2

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

Sincerely,

/7~

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, 1 0CFR 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 1 0x25 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 Page2

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 (10CFR 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 c ontainers per 10 CFR 71 to detect... "

DOCICET PROPOSED RULE

( IP ~ Fte. ~ aa *:1 --&-+) ------....:--.. DOCKETED

( (, ~ F"A. (, 3, II) us.me l"rom: "Dewhirst, Linda R." <lrdewhi@nppd.com>

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

Date: 1/8/98 6:53pm - 9 P4 :25 SUbject: Additional Comments on Final Rule 10 CFR 50.68 '98 JAN

Ms. Gallagher: OFF;cc: ri - s,r,. RY RU E) ',.,r; ~....

Recognizing the below comments are after the requested time but I A@.li~J:1) /(~,Al.

to share them anyway and ask for feedback if possible. I've heard th a t 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 10CFR50.68, 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 (b) (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!

Acknowtedg b JAN 2 1998 U.S. NUCLEAR REGULATORY ISSION RULEMAKINGS &ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics

Postmark Date ~4 ~ ~ °> 1/q/'jp, Coples Received _____ _

Add'I Reproducecl'--=-=3 -,- --- ::.--

Sp

• 1

• n f I

Linda R. Dewhirst Licensing Engineer Cooper Nuclear Station Tel: 402.825.5009 Fax: 402.825.5827 email: lrdewhi@nppd.com January 9. 1998

NOTE TO: Emile Juli an Chief. Docketing and Services Branch FROM: Carol Gallagher /). o A. o O A~

RES. ORA ~ ~0

SUBJECT:

DOCKETING OF COMMENT ON DIRECT FINAL/PROPOSED RULEMAKING

Attached for docketing is a comment letter related to the Direct

- Final/Proposed Rulemaking on Criticality Accident Requirements. This letter was received via e-mail on January 8, 1998 at 6:53 p.m. The corrvnenter's name is Linda R. Dewhirst at Cooper Nuclear Station. 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 IJfl

Northem ~.,Af Peompany

Monticello Nuclear Generating Plant 2807 West HWY..~~. Monticel5 MiQJI\\No~ 3~ ~~ i.2".'.'. ~

OOC1<ET BER,. 7 (p A 1/;~I,1r8 PROPOSED RULE !JD J 0

( "1;2. F~ ll 3B:15)

January 2, 1998 { (, ~ FR ~ 3, 11)

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 Volume 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 propo sed 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 6 2 without specific reference to that GDC. Th ird, a person unfamiliar with 10 CF R 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, ma terial selection, and geometry are major factors affecting reactivity that could compensate for highe r 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.

MONTICELLO GE 1ER TING STATION 2807 WEST HIGHWAY 75.

MONTICELLO, MN 55362-9637 Acknowledged Dy car0 JAN 2 9 1998 U.S. NUCLEAR REGUlATORY ISSION RULEMAKINGS & ADJUDICATIONS STAFF OFFICE Of THE SECRETARY Of THE COMMISSION Oocllnant S1atlstics Postmark / k, / '1 3 Coples Received _____ F I / _

- ~,

• II

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

17) ll D~

Marcus H. Voth Project Manager - Licensing

cc: T J Kim, NRR-PM, NRC Kris Sanda J E Silberg January 2. 1998

NOTE TO: Emile Juli an Chief. Docketing and Services Branch FROM: Carol Ga 11 agher RES. ORA

SUBJECT:

DOCKETING OF COMMENT ON DIRECT FINAL/PROPOSED RULEMAKING

Attached for docketing is a comment letter related to the Direct

- Final/Proposed Rulemaking on Criticality Accident Requirements. This letter was received via e-mail on January 2. 1998 at 12:52 p.m. The commenter's name and address are Marcus H. Voth. Northern States Power Company, Monticello Nuclear Generating Plant. 2807 West County Road 75. Monticello. MN 55362.

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 OOCKfTtO IISNRC

N U C L E A R E N E R G Y I N )98 T ILJM -5 p J :Q5

OF-F-1(

• AUL c, ', ' I *-. ' DI RECTOR, ENGINEERING t : f S i=(Tj 1,,: David J. Modeen ADJU t)t'<. ' :'.'t'I F F UCLEARGENERATIONDIVISION

January 2, 1998 DOCKET NLltBER Secretary PROPOSED RULE 5 O.J 1 o U.S. Nuclear Regulatory Commission ( (p :2 Ftc u, 38 :15)

Washington, DC 20555-0001 ( (,:}Fie l,3 9tt)

SUBJ E CT: 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 fluid, such as fog or sprays. Since 1976, BWR licensees have managed this concern wit h 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 licens e es an option to minimize the likelihood and impact of low-density hydrogenous fluid using administrative controls. An in-depth discussion of concern with §50.68(b)(3) is contained in Enclosure 1.

Acknowledged by car JAN 2 9 1998

1 NE! 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.

1776 I STREET NW SUITE 400 WASHINGTON DC 200D6-3708 PHONE 202 739 8000 FAX 202 785 40 19 U.S. NUCLEAR REGULATORY COMMISSION RULEMAKINGS &ADJUDICATIONS STAFF OFFICE OFlHE SECRETARY OF THE COMMISSION Ooa.ment Statisllcs PostmaJk Data Haad ]>el/ve..--ed 1/s /CJs-Coplas Recelwed ___, ___ _

Add'I Copa Reproduc8d _.,3_ ~

S~ Dtslrlbutlon -rL{ r~ ' '?D~

~ t D,s t Secretary, U.S. NRC January 2, 1998 Page 2

Neither the rule nor 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 inner 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.

David J. Modeen

KOG/edb Enclosures

c: Stan Turel, NRC/NRR S. Singh Bajwa, NRC/NRR ENCLOSURE2

ADDITIONAL COMMENTS ON THE DIB.ECT FINAL RULE

COMMENT# PARAORAPH COMMENT

1. 1 OCFR50.68(b)(1) The paragraph reads, *PJant procedures may not permit handling and transportation at any one time of more fuel assemblies than have been determined to be safely subcriticsl 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 subcritlcal" shouk:i be explicitly 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.

cl) In Paragraphs 10CFR50.68(b)(2) and 10CFR50.68(b)(4), the moderator i~ Identified as "pure wate~ rather than "unborated water,* as described in 10 CFR 50.6B(b)(1). If the moderators were 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 future misunderstandings.

e) Paragraph 10CFR50.68(b)(3) discusses "optimum moderation" by a "low-density hydrogenous fluid." The phrases "most adverse moderation* and "optimum modera1Ion* seem to express opposite relationships, but are used to describe the same physical phenomenon. When 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. 1 0CFR50.68(b)(2) The proposed paragraph reads, in part, "The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel.... * '

Since all neutrons (that are produced) subsequently either leak or are absorbed, the paragraph should be clarified to specify its applicablllty to an Instant in time. Alternately, the paragraph may be revised to eliminate the words *rat10 of neutron production to neutron absorption and leakage,* since *k--effective" is a sufficiently understood term to pennit 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 1 OCFR50.68(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, "The 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 critical 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 is not. Plant procedures and controls for SNM are adequate to control accident criticality. The paragraph should be revised to reflect this situation.

COMMENT# PARAGRAPH COMtEHT

5. 1 0CFR50.68(b)(6) The paragraph 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.*

a) 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. 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 revised to reflect the license conditions. Revise the paragraph to eliminate reference to GDC 63 and describe the underlying monitoring requirements or to require "Radiation momtors, as required by GDC 63 or other analogous licensee criteria,... "

c) The requirement that "Radiation monitors... are provided In storage and associated handling areas..." Is Inappropriate.. Fuel storage areas for both ne,,v and used fuel are not nonnally occupied volumes. As such, not all fuel storage volumes (vaults or pools) have radlatloo monitoring Inside of them. In some cases, monitoring is 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."

d) 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 to replace "initiate appropriate safety actions" with "Initiate appropriate warning."

6. 10CFR50.68(b)(7) The paragraph reads, "The maximum nominal U-235 enrichment of the fresh fuaJ assemblles Is limited to no greater than five (5.0) percent by welghe

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 crltlcality analysis for both dry and spent fuel racks to ensure the appropriate margins to crttlcality 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 KeH-<0.98 requirement of

§50.68(b)(3). These licensees would them need to submit an exemption request, unless the requirement is revised to perm.it 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 adminiRtrative 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 I

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 configuration, equivalent water densities from 0.05 to 0.45 gram/cc, the BWR fuel storage arrangement may not comply with the proposed KeH-<0.98 requirement. Licensees believe that administrative controls recommended in the SIL are appropriate to manage the Ke££ 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.

March 31, 1976 SIL No. 152 File Titb A Cat{!gory 1

CRITICALITY MARGINS FOR STORAGE OF NEW FUEL

Using optimum moderatpr conditions,. calculations 1nd1cate 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 nra extinguisher foam, water mtst, steam or other hydrogenous materials. loaded with new fuel.* Potential sources fer an optimum moderator are Th1s Service Infonnation Letter (SIL) recomnends precautionary measures to BWR operators to further reduce the already very low probability of such an event occurring.

DISCUSSION An.analysis by General Electric indicated that it would require the introduction of a low et(Uivalent water density material to completely occupy.. the space in.and around an.array of fwel - assemblies 1n storar~

ar_rays were analyzed. with and without gadolinia, to simulate reactivity for the occurrence of.a criti~ality. Both Io x 25 and 20 x* 25 bund e condftions from initi"al-core loads* to the most reactive ctesign basis equi-valent water ~s1.ty was

• 0.2 gram/cc reload fuel. In all cases the optimum,noder.at*fon occurred when the *. In the worst ca~e. a range of equivalent water densities from.Q~OS to 0.45 irams/cc was undesirable in confonning to the design basis* Keff limits *. This concern Has been judged by General ElectM~ not to be a reportable deficiency and the judgJr.ent has been supported by the NRC.

statfons and sprinklers on the refueling flonr at a significant Resul~~ of BWR ~ite sµrveys ha~e indicated the presence.of ffre hose number of plants. Also, a substantial number of hoses in these a solid stream to a coarse spray. In the interest of assuring safety stations are provided with adjustable noz=les that are variable rrom margins on the refueling floor. additional controls that further reduce the probability of a criticality occurrence should be implemented.

RECOMMENDED ACTION General Electric recorrmends that the proceduroT controls listed below be c;ons1dered at the earliest convenient opportunity to reduce the for new fue remote probability for inadvettently establishing critica1*conditions 1 s tora-ge-: * * * * :. * *.. *

• NUCLE.\\R ENERGY DIVISION _;. OPERA~G PLANT SERVICES

• SAN JOSE, CALtFORNIA 65125 NO WARl'tAHTI' 0a REPRESl!NTATlOlll ElCJIIIIIEssm.OR IMPU!D nt MADE WITH

RISP&CT TO THE ACCUR-.CV

• COMPt.E'J'ENESfl OR USll'ULNE$ OP THIS INFDR*

M.-T1011#. GENEAAt. ELECTRIC COMPANY MSUM'SS NO RESPONSIBILITY POA LIA* ~

• r CT~ I r, BILITY OR DAMAGc WHICM MA y Rest.lLT FROM THE USE OF THIS INPORM* TION. G E ti ER A L C;: ~ ~ l 1:: *,.

• March 311 1976 SIL No. 152 Attachment

PROCEDURAL RECOMMEtjDATiotlS FOR NORMAL FUEL HANOLrnG OPERA~IONS

1. No more than one fuel bundle should be suspended above the fuel storage array and this at a height no greater than 24 1nd1es to lfmit penet~ation displacement if the bundle was dro?ped.

2. Fuel handling in the fuel storage area should be limited tD 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 overload test weight should at no time be suspended abov~ the fuel storage array.

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

• I *

4. A fuel storage areas or properly d~signed fuel shipping container fuel array of four or more fuel bundles outside of the nonna1 should be prohibited.

5. The new fuel vault should always be dry.

6. The spent fuel pool ~hou1d be maintained in a-flooded condition construction debris are pre~ent. Flooding ~hould prov1d~ ~t le~st if new fuel is in storage during construction activities or: when

• enough not flooded when new fuel is in storage~ the fuel should be water to cover the bundles. If the spent fuel pool is covered by a solid, firepr.oof material to prevent possible 1nundat1on 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 ff abnonnal condition of vault flooding occurs. an

9. Fuel should not be placea in aisles or moved through aisles adjacent to and at the same l~vel 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 is stored in temporary star-age racks below the fuel pool..

work table. the work table.should not be used to handle fuel; conversely, 1f the'work table is used to handle fuel, fuel storage below the work table should be prohibited..

March 26. 1976 SIL No. 152 Attachment

12. No more than two fuel bundles should be allowed in or around a fuel prep machine at any ~fme. This fuel should be separated frocn the main body of stored_f:uel by at least 12 inches.

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

14. New fuel should not be stored in the nN fuel vault when there are construction activities on the refueling floor or construction debris 1n the vicinity of the new fuel 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 example, it should be i~ossible to block the drain, or in.any:way produce the equivalent water densities in the ranges noted above.

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

3. Fuel should not be stored in the new fuel vault when there are construction activities on the refueling floor or construction debris in the vicinity of the new fuel vault unless a solid cover is placed over the vault. This solid cover would as a fog or spray should the operation of fire hoses become help to prevent the introduction of low density water such necessary on the refueling floor.

4. The attachment to this SIL*entitled. "Procedural Reconmendations for Normal Fuel Handling Operations" should be reviewed by

&"!R plant personnel to assure a complete understanding of all current procedura1 controls relevant to fuel handling oper~tions. It should be noted that for reasons of emphasis items 5, 6 and.. !4 ;n the attachment are identical to the procedural control recoorriendations 1. 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 ~. iz;; Issued by: anager I)~[. Layt~anager Product Service Service Cacnnunicat1ons

Product

Reference:

A71 - Plant Recanmendnt1ons Jll - fuel and Reloads JAN. 2.1998 2 : 45PM SOUTHERN NUCLEAR (205)992 6108 N0.025 P.2/3

Lawis S11mn1r Southam Nuclear Vice President Dp,miq Cainpaay, Inc.

Hat-ch Project Support 40 Inverness Partway DOCKE TED Post Office Box 1295 USNRC Birmingham, Alabama 35201 Tel 205.992.7279 Fax 2.05 992.0341

December 31, 1997

Docket Nos. SQ.. 321 50-348 50-424 HL-5546 50-366 50-364 50-42S LCV.. 1145

Mr. John C. Hoyle Secretary of the Commission DOCKET NUMBER U.S. Nuclear Regulatory Commission PROPOSED RULE 5 0.J-7 0 Washington, DC 20SSS-0001 ( ~,R FR t,39,25).

A ITENTION: Rulemaking and Adjudications Staff ( t, ~ F/2 ~ 3 '1 II)

Comments on Direct Final Rule "Criticality Accident Requirements" (62 Federal Register 63825 dated December 3, 1997)

Dear Mr. Hoyle:

On December 3, 1997, the Nuclear Regulatory Comn;rission (NRC) published in the Federal Register concurrently 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 50.68 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, please advise.

Respectfully submitted1

H. L. Sumner, Jr.

HLS/TMM

( distribution - see next page)

_...-4 JAN 2 9 1998 Acknowledged by \\;QIU r r er **

U.S. NUCt.EAR REGULATORY COM ISSION RULEMAKINGS & ADJUDICATIONS St OFFICE OF lliE SECRETARY OFlliE COMMISSION Docllnent Statistics Postmartc oate Jvt ut QY), I~ /. Lg Coples Racetved ___..___ _ ___.....,

Add'I Copies Reproduced.3 _..

SpecialDislrtbution -Tl.Ar~(. ?'D & _

'1?1DS I JAN. 2.1998 2 =46PM SOUTHERN NUCLEAR (205)992 6108 N0.025 P.3/3

U. S. Nuclear Regulatory Commission Page2

cc: Southern NuQleN' Operatini Compan,v Mr. D. N. Morey, Vice President., Plant Farley Mr. C. K. McCoy, Vice President, Plant Vogtle Mr. J.B. Beasleyt General Manager - Plant Vogtle

Mr. R. D. Hill, General Manager* Plant Farley Mr. P. H. Wells, General Manager - Plant Hatch u, s, Nuclear Rc:iJ1latocy Commission,JVashln~n. nc 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, Nuclearlkiulmcy Commission, Relliou 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-5546 LCV-1145 CP&L

Carolina Power & Light Company PO Box 1551 4 11 Fayetteville Street Mall CP&L Letter: PE&RAS-97-101 Raleigh NC 27602 December 24, 199 7

DOCKET NUMBER PRO OSED RULE S o+ 70

( IP;lF~(,~8~ 5 )

( (, :2F~ ~3etlf)

- Secretary U.S. Nucle ar 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 Performance Evaluation & Regulatory Affairs

HAS Attachment cknowiedged by DEC ~ 1 \\997 U.S. NUCLEAR REGULATOR Y COMMISSION RULEMAKINGS & ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION Document Statlslics Postmark Date,~1~~fe;1,- 1?.tce;tel f'r f)rn lt4rol G1.lkj~ey 011,~!~fft;~ w~ t-tfr,w,tl c(tx11,,.~,.f l'ron-i Copies Recavad / 7 1q-ftr1iefi "~ rlA,~~,._j,-"'lj Uthsif~

Add'I Copies Reproduc:E}_g __ 3...,...-:--=--~

SpecialDistnbution,_ ~ r.~t~.i...:::-.~:.:..u::::-

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-HBR SEP, 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 Criti ca lity Accident Requirements (62 FR 63825 and 62 FR 6391 1)

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 I 0CFR50.68(b )(2) reads, in part:

"The estimated ratio of neutron production to neutron absorption and leakage (k-ejfective) 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 wo rds "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 10CFR50.68(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 of the fresh fuel assemblies is limited to no greater than five (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 Julian Chief. Docketing an d Services Branch J FROM: Carol Gallagher /1_.P. ~

RES. ORA ~

SUBJECT:

DOCKETING OF COMMENT ON DIRECT FINAL/PROPOSED RULEMAKING

Attached for docketing is a comment letter related to th e 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 Commonwealth Edison Company 1400 Opus Place Downers Grove, IL 60515-5701 (!)

DOCK ETE D USNRC

'97 DEC 29 P 2 :53 ComEd

December 22, 1997

U. S. Nuclear Regulatory Commission DOCKET NUMBER ATTN : Rulemaking and Adjudications Staff PROPOSED RULE 5 O J-10 Washington, DC 20555-0001 ( ~ a. FI< <, 38~5)

( " ;2 F~ G, 3 't I!)

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.

l ~

T homas J. Ko v ach /:-

Vice P resident Nuclear Regulatory Services

Attachment

cc: G. Dick, Generic Issues Project Manager - NRR A. B. Beach, Regional Administrator - RIii Office of Nuclear Safety - IONS

Ackrlowfedged by' card e w.. DEC 3 1 1997.._............ 888118fll iJifMo

A Unicom Company U.S. NUCLEAR REGULATORV COMMtSSIOO AULEMAKINGS & ADJUDICATIONS STAFF OFACE OF THE SECRETARY OF THE COMMISSION Docllnent Statistics Postmar1< Date I 2 /zi /1; 1 Copies Received __, _ ___ _

Add'I Copies Reproduced---~

Special 1? I 'fl$ r u., Distribution 1'1 r~ l "?l> R 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 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 IO 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 10CFR70.14. 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 following requirements in lieu of maintaining a monitoring system capable of detecting a criticality as described in JO 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.

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 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 /ow-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.

Comments:

For the ComEd Boiling Water Reactors (BWRs), optimum mo deration 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, 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 percent confidence level, if flooded 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 1 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 DOCKET NlMBER..

PROPOSED RULE p 5 0 f 7 0 DO CKETED

(<,'J.FR.t,38'~S) USNRC (7590-01-P]

NUCLEAR REGULA TORY COMMISSION "97 OEC -9 A10 :35

10 CFR Parts 50 and 70

RIN: 3150-AF87

Criticality Accident Requirements

AGENCY: Nuclear Regulatory Commission.

ACTION: Direct Final rule with Opportunity to Comment.

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,

- us ed, 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

ne gligible.

1~ I '1 1 l'i'il' EFFECTIVE DATE: The final rule is effective (~5 day& after 191:tbhealioA iA &l:le Feder.al Register),

1 unless significant adverse comments are received by (30 days ~=l,';*n.~\\n~F?fera1

Begii.ier). If the effective date is delayed, timely notice will be published in the Federal

Register.

ADDRESSES: Mall comments to: Secretary, U.S. Nuclear Regulatory Co~mlssion,

Washington, DC 20555-0001, Attention: Rulemaking and Adjudications Staff.

Hand deliver comments to 11555 Rockville Pike, Maryland, between 7:30 am and

4:15 pm on Federal workdays.

Coples 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, Washington, DC 20555-0001, telephone (301)

415-6234, e-mail spt@nrc.gov.

SUPPLEMENTARY INFORMATION:

Background

The Nuclear Regulatory Commission (NRC) is 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 certain requirements that would

be incorporated into 10 CFR Part 50. These are generally the requirements that the NRC has

used to grant specific exemptions to the requirements of 10 CFR 70.24. In addition, the NRC is

revising the current text of the section relating to seeking specific exemptions from regulations

in 10 CFR 70.24(d) which provided that a licensee could seek an exemption to all or part of

- 2 -

10 CFR 70.24 for good cause because it is redundant to 10 CFR 70.14(a). A modified 10 CFR

70.24{d) ~s being added to provide that the requirements in paragraph (a) through {c) of 10 CFR

Part 74 do not apply to holders of a construction permit or operating license for a nuclear power

reactor issued pursuant to 1 0 CFR Part 50, or combined licenses issued under 1 0 CFR Part 52,

if the holders comply with the requirements of 10 CFR 50.68 {b).

The Commission's regulations In 10 CFR 70.24 require that ea<?fl licensee authorized to

possess more than a small amount of SNM maintain a criticality monitoring system_"uslng

• gamma-or neutron-sensitive radiation dt;ttectors which will energize clearly audible alarm

signals if accldental criticality occurs" In each area in which such material is handled, used, or

stored. The regulation also specifies sensitivity requirements for these monitors and details the

training that licensees must conduct in connection with criticality monitor alarms. The purpose

of this section is to ensure that if a criticality were to occur during the handling of SNM,

personnel would be alerted and would take appropriate action.

Most nuclear power pla'nt licensees were granted exemptions from 10 'CFR 70.24 during

the construction of their plants as part of the 10 CFR Part 70 license issued to permit the

receipt of the initial core. Generally, these exemptions were not explicitly renewed when the 1 0

CFR Part 50 operating license, which now contained the combined Part 50 and Part 70

authority, was Issued. The requirements in 10 CFR 70.24 prescribe the attributes required of

the monitoring and alarm system. Compliance with these requirements may be unnecessary

for commercial power reactors where the conditions which could lead to a criticality event are

so unlikely that the probability of occurrence of an Inadvertent criticality is negligible. The NRC

anticipated that the regulation might be unnecessary for some licensees and included In 1 0 I

CFR "?0.24(d) an Invitation to any licensee to seek an exemption to the entire section or part of

the section for good cause. A large number of exemption requests have been submitted by

. - 3 -

power reactor licensees and approved by the NRC based on safety assessments which

concluded that the likelihood of criticality was negligible. Because of th~ experience gained in

processing these exemption requests, the NRC concluded that the regulations should be

amended to provide this flexibility without requiring licensees to-go through the exemption

process.

Discussion

At a commercial nuclear power plant, the reactor core, the fresh fuel delivery area, the

fresh fuel storage area, fhe spent fuel pool, and the transit areas among these, are areas where I

amounts of SNM sufficient to cause a criticality exist. In addition, SNM may be found In

laboratory and storage locations of these plants, but an Inadvertent criticality is not considered r

credible in these areas due to the amount and configuration of the SNM. The SNM that could

be assembled into a critical mass at a commercial nuclear power plant is only in the form of

nuclear fuel. Nuclear power plant licensees have procedures a_nd the plants have design

features to prevent inadvertent criticality. The inadvertent criticality that 10 CFR 70.24 is

intended to address could only occur during fuel-handling operations.

In contrast, at fuel fabrication facilities SNM is found and handled routinely in various

configurations in addition to fuel. Although the handling of SNM at these facilities is controlled

by procedures, 'the variety of forms of SNM and the frequency with which it is handled provide

greater opportunity for an inadvertent criticality than at a nuclear power reactor.

At power reactor facilities with uranium fuel enriched to no greater than five (5.0) percent

by weight U-235, the SNM in the *fuel assemblies cannot go critical without both a critical

configuration and the presence of a moderator. Further, the fresh fuel storage array and the

spent fuel pool are in most cases designed to prevent inadvertent criticality, even in the

- 4 -

presence of an optimal density of unborated moderator. Inadvertent criticality during fuel

handling is precluded by limitations on the number of fuel assemblies permitted out of storage

at the same time. In addition, General Design Criterion (GDC) 62 in Appendix A to 1 0 CFR Part

50 reinforces the prevention of criticality in fuel storage and handling through physical systems,

processes, and safe geometrical configuration. Moreover, fuel handling at power reactor

facilities occurs only under strict procedural control. Therefore, the NRC considers a fuel

handling accidental criticality at a commercial nuclear power plant to be extremely unlikely. The

NRC believes the criticality monitoring requirements of 1 0 CFR 70.24 are unnecessary as long

as design and administrative controls are maintained.

Because the NRC considers an inadvertent criticality to be unlikely at a nuclear power

reactor, by this rulemaking it is granting nuclear power reactor licensees a choice - either meet

the criticality monitoring requirements of 10 CFR 70.24 or in lieu of those criticality monitoring

requirements meet certain criteria related to procedures, plant design, and fuel enrichment.

These criteria are incorporated into section 50.68(b) of 10 CFR Part 50 by this direct final rule.

The three changes in the requirements are as follows:

(1} Section 50.68(a) provides that each holder of a construction permit or operating license for

a nuclear power reactor issued under Part 50, or a combined license for a nuclear power

reactor issued under Part 52 shall comply with either 10 CFR 70.24 or the seven requirements

in section 50.68(b).

(2) Section 50.68(b) provides that each licensee as described in 50.68(a) shall comply with the

seven listed requirements in lieu of maintaining a monitoring system capable of detecting a

criticality as described in 10 CFR 70.24.

(3) The revised section 70.24(d) provides that the requirements in 10 CFR 70.24(a) through (c)

do not apply to holders of a construction permit or operating license for a nuclear power reactor

- 5 -

issued pursuant to 10 GFR Part 50, or combined licenses issued under 10 GFR Part 52, if the

holders comply with the requirements of paragraph (b) of 10 GFR 50.68.

Procedural Background

Because NRG considers these amendments to its rules to be noncontroversial and

routine, public comment on these amendments is unnecessary. The amendments to the rules

will become effective on (75 days after publication in the Federal Register). However, if the

NRG rec::eives significant adverse comments on the companion proposal published concurrently

in the proposed rules section of this Federal Register by (30 days after publication in the

Federal Register), then the NRG will publish a document that withdraws this action and will

address the comments received in response to the amendments. Such comments will be

addressed in a subsequent final rule. The NRG will not initiate a second comment period on

this action.

Findings

Upon review of this rulemaking, that the changes and additions addressed by this

rulemaking do not significantly affect the environmental cost-benefit balance that otherwise

would justify the licensing of a light-water nuclear power reactor. The basis for this finding is

that this rule is a codification of practices in place and does not significantly affect the cost

benefit balance for a light-water reactor.

- 6 -

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.

Environmental Impact: Categorical Exclusion

The NRC has determined that this proposed regulation is the type of action described in

categorical exclusion 10 CFR 51.22(c)(3). Therefore neither an environmental impact

statement nor an environmental assessment has been prepared for this proposed regulation.

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 r

as files (any format), if your web browser supports that function. For information about the I

interactive rulemaking site, contact Ms. Carol Gallagher, (301) 415-6215; e-mail CAG@nrc.gov.

Paperwork Reduction Act Statement

This direct final rule does not contain a new or amended information collection requiremert

subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). Existing requirements

- 7 -

were approved by the Office of Management and Budget, approval numbers 3150-0009 and 3150-

0011.

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.

Regulatory Analysis

The structure of the current 10 CFR 70.24 is overly broad and places 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 the potential to result in a large number of recurring exemption requests.

To relieve the burden on power reactor licensees of applying for, and the burden on the staff

of granting recurring exemptions, this amendment permits power reactor facilities with fuel

enrichments no greater than 5 weight percent U-235 to be, excluded from the scope of 10 CFR

70.24, provided they meet specific requirements being added to 10 CFR Part 50. This a'mendmert

fs a result of the experience gained in processing and evaluatlrg a number of exemption req~ests

from power reactor licensees and NRC's safety assessments in response to these requests that

concluded that the likelihood of criticality was negligible.

11"1e only other viable option to this amendment is for the NRC to do nothing and allow the licensees to continue requesting,exemptlons. If nothing Is done, the llcensees ' will continue to incur

the costs of submitting exemptions and NRC will Incur the costs of reviewing them. Under this rule,

- 8 -

an easing of burden on the part of licensees results by their not having to request exemptions.

Similarty, the NRC wm not need to review and evaluate these exemption requests, r:esulting in an

easing of burden for the NRC.

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 licensees, has no

negative impact on the public health and safety, but will provide certain licensees savings, and

savings to the NRC as well. Hence, the rule is shown to be cost beneficial.

The foregoing constitutes the regulatory analysis for this final rule.

Regulatory Flexibility Certification

In accordance with the Regulatory Flexibility Act of 1980, 5 U.S. C. 605(b ), the Comrnissbn 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 licensees, 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).

Backfit Analysis

The Commission has determined that a backfit analysis is not needed. This rule Is a

cocfdication of practices In place by the NRC and is not a modification of or addition to systems,

structures, components, or desig[I of a facility; or the design approval or manufacturing license for

a facility; or the procedures of organization required to design, construct or operate a facility; any

- 9 -

of which may result from a new or amended provision in the Commission rules or the imposition

of a regulatory staff position Interpreting the Commission rules that is either new or different from a previously applicable NRC staff position (1 O CFR Chapter I).

Small *Business Regulatory Enforcement Fairness Act

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

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, Reporti~

and recordkeeping requirements.

• 10 CFR Part 70

Criminal penalties, Hazardous materials transportation, Material control and accounting,

Nuclear materials, Packaging and containers; Radiatio11 prot~ctlon, ~epc;>rting and record keeping

requirements, Scientific equipment, Security measu*res, Special nuclear material.

For the reasons set out in the preamble and under the authprtty of the Atomic Energy Act

of 1954, as amended, the Energy Reorganization Act of 1974, as amended, the National

- 10 -

Environmental Policy Act of 1969, as amended, and 5 U.S.C. 553, the NRC is adopting the

following amendments to 1 O 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, as amended, 202, 206, 88 Stat.

1242, as amended 1244, 1_246, (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). 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. 21 ~8). Se.ctions 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 Appe~dix 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 LI.S.C. 2239}. Section 50.78 also issued under

sec. 122, 68 Stat. 939 (42 U.S.C. 2152). Sections50.80 50.81 also issued under sec. 184, 68 Stal

954, as amended (42 U.S.C. 2234). Appendix Falso issued under sec. 187, 68 Stat. 955 (42 u.s.c. 2237).

- 11 -

2. Section 50.68 is added under the center heading "Issuance, Limitations, and Conditions

of Licenses and Construction Permits" to read as follows:

(

§ 50.68 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 n,uclear 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 r~quirements 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 3ssemblies 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

I 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 enrichm~nt 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

- 12 -

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

(subcrittcal), at a 95 percent probability, 95 percent 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

hanc:lfing 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 ~ad 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 I

amended, 1244, 1245, 1246, (42 U.S.C. 5841, 5842, 5845, 5846).

Sections70.1(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 (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).

- 13 -

.:.ction 70.61 al ed under secs. 186, 187, 68 S~ (42 U.S.C. 2236, 2237).

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

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

§ 70.24 Qriticanty accident requirements.

(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.

-ii,..

Dated at Rockville, Maryland this lY day of N 6V, 1997.

For the Nuclear Regulatory Commission.

~

Executive Director ~or Operations.

\\

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