Public Service Electric and Gas Co.; Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-272 and 50-311]

Public Service Electric and Gas Co.; Consideration of Issuance of

Amendment to Facility Operating License, Proposed No Significant

Hazards Consideration Determination, and Opportunity for a Hearing

The U.S. Nuclear Regulatory Commission (the Commission) is

considering issuance of an amendment to Facility Operating License Nos.

DPR-70 and DPR-75 issued to the Public Service Electric and Gas Company

(the licensee) for operation of the Salem Nuclear Generating Station,

Units 1 and 2, located in Salem County, New Jersey.

The proposed amendment would increase the storage capacity of the

spent fuel pool (SFP) from its current 1170 storage cells to 1632

storage cells. This would be accomplished by replacing 9 out of twelve

of the existing high density fuel racks with 9 maximum density rack

modules constructed of stainless steel and a neutron absorber material

(boron carbide and aluminum-composite sandwich, product name

``boral''). The proposed change would extend the date when full core

discharge capacity is no longer available for Salem 1 from 1998 to

2008, and for Salem 2 from 2002 to 2012.

In addition, the proposed amendment would extend the decay time for

refueling operations from 100 hours to 168 hours.

Before issuance of the proposed license amendment, the Commission

will have made findings required by the Atomic Energy Act of 1954, as

amended (the Act) and the Commission's regulations.

The Commission has made a proposed determination that the amendment

request involves no significant hazards consideration. Under the

Commission's regulations in 10 CFR 50.92, this means that operation of

the facility in accordance with the proposed amendment would not (1)

involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

or (3) involve a significant reduction in a margin of safety. As

required by 10 CFR 50.91(a), the licensee has provided its analysis of

the issue of no significant hazards consideration, which is presented

below:

1. Do not involve a significant increase in the probability or

consequences of an accident previously evaluated.

[Public Service Electric and Gas Company] PSE&G has evaluated

the following postulated accident scenarios:

1. A spent fuel assembly drop in the SFP.

2. Loss of SFP cooling.

3. A seismic event.

4. An installation accident during reracking.

The Salem SFP has been analyzed considering fuel handling

equipment, operating procedures, SFP cooling system, and seismic

events. Reracking involves replacing 9 out of the 12 existing high

density racks with 9 new maximum density racks. It does not require

any system modifications or modifications to the cask handling

crane, which by its physical location and design is prevented from

moving over the SFP. Results confirm that the proposed modification

does not increase the probability of the first three postulated

accident scenarios.

NUREG-0612, ``Control of Heavy Loads at Nuclear Power Plants,''

sections 5.1.1, 5.1.2, and 5.1.6, provide guidance for heavy load

handling operations during spent fuel storage rack replacement.

Section 5.1.2 lists (4) alternatives for assuring safe heavy load

handling during a fuel storage rack replacement. Alternative (1)

satisfies the control of heavy loads guidelines through the

implementation of defense-in-depth measures. These measures ensure

that the potential for a heavy load drop is extremely small. PSE&G

intends to utilize the defense-in-depth concept during reracking

activities.

NUREG-0554, ``Single Failure Proof Cranes for Nuclear Power

Plants,'' provides guidance for the design, fabrication,

installation, and testing of highly reliable new cranes. NUREG-0612,

Appendix C, ``Modification of Existing Cranes,'' provides guidance

for the implementation of NUREG-0554 at operating plants. We have

evaluated anticipated fuel handling crane movements for compliance

with the guidelines specified in alternative (1) of Appendix C, and

determined that alternative (1) was satisfied based on the extremely

small probability of a storage rack drop. The maximum weight of any

storage rack and its associated handling tool is 17 tons. The fuel

handling crane will be upgraded to a 20 ton lifting capacity and a

design safety factor, with respect to ultimate strength, of five

times the lifting capacity (i.e., 100 tons). The uprated fuel

handling crane has ample safety factor margin for storage rack

movement. This applies to non-redundant load-bearing components.

Special redundant lifting devices, which have a rated capacity

sufficient to maintain safety factors, will be utilized for storage

rack movements. Per NUREG-0612, Appendix B, the substantial safety

factor margin ensures that the probability of a load drop is

extremely low. Additionally, a load drop analysis was performed to

ensure the integrity of the pool structure. The analysis results

were acceptable.

Based on the actions discussed above, the proposed modification

does not increase the probability of an installation accident.

PSE&G evaluated the consequences of a spent fuel assembly drop

in the SFP and determined that the criticality acceptance criterion,

Keff less than or equal to 0.95, was not exceeded. The radiological

consequences of a fuel assembly drop did not change significantly

from those previously analyzed. The calculated doses are well within

10 CFR 100 requirements. A spent fuel assembly dropped on the racks,

will not cause rack distortion that would prevent the performance of

their safety function. Thus, the consequences of this postulated

accident are not significantly changed from those previously

evaluated.

The consequences of a loss of SFP cooling were evaluated. The

evaluation concluded that sufficient time is available to establish

an alternate means of cooling following a complete failure of the

normal SFP cooling system. Calculations show that under a normal

discharge scenario, if all indirect forced cooling paths (i.e., heat

removal by heat exchangers) are lost at the instant the pool water

reaches its maximum value, the pool will not begin bulk boiling for

at least 4.61 hours. This time interval is sufficient to allow plant

personnel to establish alternate heat removal methods. A piped

cross-connection exists between Unit 1 and Unit 2's SFP heat

exchangers. This allows for use of the opposite Unit's heat

exchanger during emergencies, or when a given Unit's Service Water

header or Component Cooling System are out-of-service. Thus, the

consequences of this postulated accident are not significantly

changed from those previously evaluated.

The new racks are designed and fabricated to meet applicable NRC

requirements and industry standards. Seismic analyses were performed

on the new racks and the existing racks using 3-D single rack

(opposed phase motion) and Whole Pool Multi-Rack (WPMR) models.

Kinematic and shear analyses conclude the existence of large margins

of safety. The kinematic margin against rack-to-rack or rack-to-wall

impact is at least 1.5 for all SFP racks. Maximum rack primary

stresses, under [Safe Shutdown Earthquake] SSE conditions, are less

than 50% of the allowable ASME Code value. Maximum supporting pool

structure bending moments and thru-thickness shear, under factored

load conditions, are less than 80% of the allowables. All racks (new

and existing) are designed as free-standing racks, to ensure that

rack and pool structure integrity is maintained during and after a

seismic event. Thus, the consequences of a postulated seismic event

are not increased from previously evaluated events.

The consequences of an installation accident were considered.

All fuel in the SFP will have decayed for a minimum of (3) months

prior to any heavy load movement in the SFP area. This allows

sufficient time for decay of gaseous radionuclides in the fuel (gap

activity). A postulated accidental gaseous release from all stored

fuel assemblies would result in a potential offsite dose less than

10% of 10CFR100 limit. No equipment essential to safe reactor

shutdown or employed to mitigate the consequences of an accident is

located beneath, adjacent to, or within the area of influence of any

load handling to support the SFP modification. Thus, the

consequences of a postulated installation accident are not

significantly increased from those previously evaluated.

The only postulated accident affected by decay time is a Loss of

SFP cooling. The proposed increase in decay time prior to refueling

operations is conservative and decreases the decay heat removal

requirements. All thermal-hydraulic calculations used 168 hours as

the assumed decay time and concluded that adequate heat removal

capability existed. Thus, the probability and consequences of a loss

of SFP cooling accident are not significantly increased from those

previously evaluated.

Therefore, it may be concluded that the proposed changes do not

increase the probability or consequences of an accident previously

evaluated.

2. Do not create the possibility of a new or different kind of

accident from any previously evaluated.

The proposed modification has been reviewed and analyzed for

possible accidents. The criteria used in the analyses, design, and

installation of the new spent fuel racks account for anticipated

loadings and postulated conditions that may be imposed upon the

structure during its lifetime, and is in conformance with

established codes, standards, and specifications acceptable to the

NRC.

Factors that could affect the SFP neutron multiplication factor

have been addressed conservatively. PSE&G concluded that the maximum

SFP neutron multiplication, with the addition of the maximum density

racks, will not exceed the subcritically limit of Keff less than or

equal to 0.95.

The increase in decay time prior to refueling operations reduces

the initial heat load and SFP cooling equipments. The addition of

new racks and associated spent fuel will produce an incremental heat

load in the SFP. However, analysis has shown that the existing SFP

cooling system is sufficient to absorb this incremental heat load.

The peak bulk pool temperature will be maintained below the

threshold value to preclude bulk boiling. The incremental heat load

does not alter SFP cooling safety considerations from those

previously reviewed and found acceptable.

Rack impact analysis was performed to investigate possible

impact during seismic events (i.e., rack-to-rack and rack-to-wall

impacts). The analysis concluded that the proposed SFP modification

does not result in rack-to-rack impact in the cellular region or

rack-to-wall impact during postulated seismic events.

The basic SFP reracking technology has been reviewed and

approved by the NRC in numerous applications for spent fuel capacity

increases. The safety function and operation of the SFP cooling

system, makeup, and structural systems are unchanged by the

modification. No new failure modes are created.

Therefore, it may be concluded that the proposed changes do not

create the possibility of a new or different kind of accident from

any previously evaluated.

3. Do not involve a significant reduction in a margin of safety.

The safety function of the SFP and the racks is to preclude

inadvertent criticality in a safe, specifically designed, underwater

storage location for spent fuel assemblies that require shielding

and cooling during storage and handling. The NRC Staff has

established that the issue of margin of safety, when applied to

reracking modifications, should address the following areas:

1. Nuclear criticality considerations.

2. Thermal-hydraulic considerations.

3. Mechanical, material, and structural considerations.

Assessment in these areas assures that the SFP and racks will

withstand specified design conditions, without impairment of the

structural integrity or performance of required safety functions.

The criticality analysis confirms that the new and existing rack

designs meet the NRC acceptance criterion of Keff less than or equal

to 0.95 under all conditions. The criticality analysis methods

conform to applicable industry codes, standards, specifications and

NRC guidance.

Keff calculations include uncertainties at a 95%/95% probability

confidence level. Thus, the proposed amendment does not involve a

significant reduction in the nuclear criticality margin of safety.

Conservative methods and assumptions were used to calculate the

maximum fuel temperature and the increase in SFP water temperature.

The thermal-hydraulic evaluation employed methods previously used to

evaluate existing spent fuel racks. The results demonstrate that the

temperature margins of safety are maintained. The proposed

modification, with the fuel inventory, will increase the heat load

in the SFP. However, the decay time prior to refueling operations

was increased from 100 to 168 hours to reduce the initial SFP

cooling requirements. Evaluation results indicate that the existing

SFP cooling system can maintain the bulk pool water temperature at

or below 149 F under normal discharge scenarios. The maximum

allowable temperature for bulk boiling is not exceeded for the

calculated increase in pool heat load. Maximum local water

temperatures, along the hottest fuel assembly, remain below the

nucleate boiling condition. While no nucleate boiling is indicated

for the standard storage condition, an assumption of 50% cell

blockage results in a possible highly localized two-phase condition

near the top of the fuel. Fuel clad thermal stresses remain less

than 7000 psi, which is considerably lower than the endurance limit

of the clad material. Thus, there is no significant reduction in the

margin of safety for thermal-hydraulic or SFP cooling.

Maintaining the spent fuel assemblies in a safe configuration

during normal and abnormal loadings is the primary safety function

of the SFP and racks. Abnormal loading associated with an

earthquake, a spent fuel assembly drop, or the drop of any other

heavy object were considered. The mechanical, material, and

structural design of the new spent fuel racks complies with

applicable portions of the NRC OT Position Paper. Rack materials are

compatible with the spent fuel pool environment and the spent fuel

assemblies. The structural assessment of the new racks concluded

that tilting and deflection or movement will not result in impact in

the active fuel region during postulated seismic events. In

addition, the spent fuel assemblies remain intact with no

criticality concerns. Thus, there is no significant reduction in the

margin of safety for mechanical, material and structural

considerations.

Therefore, it may be concluded that the proposed changes do not

involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

The Commission is seeking public comments on this proposed

determination. Any comments received within 30 days after the date of

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of the 30-day notice period. However, should circumstances

change during the notice period such that failure to act in a timely

way would result, for example, in derating or shutdown of the facility,

the Commission may issue the license amendment before the expiration of

the 30-day notice period, provided that its final determination is that

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received.

Should the Commission take this action, it will publish in the Federal

Register a notice of issuance and provide for opportunity for a hearing

after issuance. The Commission expects that the need to take this

action will occur very infrequently.

Written comments may be submitted by mail to the Rules Review and

Directives Branch, Division of Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555, and should cite the publication date and page

number of this Federal Register notice. Written comments may also be

delivered to room P-223, Phillips Building, 7920 Norfolk Avenue,

Bethesda, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.

Copies of written comments received may be examined at the NRC Public

Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC

20555.

The filing of requests for hearing and petitions for leave to

intervene is discussed below.

By April 4, 1994, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and any person whose interest may be affected by this

proceeding and who wishes to participate as a party in the proceeding

must file a written request for a hearing and a petition for leave to

intervene. Requests for a hearing and a petition for leave to intervene

shall be filed in accordance with the Commission's ``Rules of Practice

for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested

persons should consult a current copy of 10 CFR 2.714 which is

available at the Commission's Public Document Room, the Gelman

Building, 2120 L Street, NW., Washington, DC 20555 and at the local

public document room located at the Salem Free Public Library, 112 West

Broadway, Salem, New Jersey 08079. If a request for a hearing or

petition for leave to intervene is filed by the above date, the

Commission or an Atomic Safety and Licensing Board, designated by the

Commission or by the Chairman of the Atomic Safety and Licensing Board

Panel, will rule on the request and/or petition; and the Secretary or

the designated Atomic Safety and Licensing Board will issue a notice of

hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene

shall set forth with particularity the interest of the petitioner in

the proceeding, and how that interest may be affected by the results of

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

following factors: (1) The nature of the petitioner's right under the

Act to be made party to the proceeding; (2) the nature and extent of

the petitioner's property, financial, or other interest in the

proceeding; and (3) the possible effect of any order which may be

entered in the proceeding on the petitioner's interest. The petition

should also identify the specific aspect(s) of the subject matter of

the proceeding as to which petitioner wishes to intervene. Any person

who has filed a petition for leave to intervene or who has been

admitted as a party may amend the petition without requesting leave of

the Board up to 15 days prior to the first prehearing conference

scheduled in the proceeding, but such an amended petition must satisfy

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

scheduled in the proceeding, a petitioner shall file a supplement to

the petition to intervene which must include a list of the contentions

which are sought to be litigated in the matter. Each contention must

consist of a specific statement of the issue of law or fact to be

raised or controverted. In addition, the petitioner shall provide a

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

contention and on which the petitioner intends to rely in proving the

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

aware and on which the petitioner intends to rely to establish those

facts or expert opinion. Petitioner must provide sufficient information

to show that a genuine dispute exists with the applicant on a material

issue of law or fact. Contentions shall be limited to matters within

the scope of the amendment under consideration. The contention must be

one which, if proven, would entitle the petitioner to relief. A

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses.

If a hearing is requested, the Commission will make a final

determination on the issue of no significant hazards consideration. The

final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

request for a hearing. Any hearing held would take place after issuance

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before issuance of any amendment.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555, Attention: Docketing and Services

Branch, or may be delivered to the Commission's Public Document Room,

the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the

above date. Where petitions are filed during the last 10 days of the

notice period, it is requested that the petitioner promptly so inform

the Commission by a toll-free telephone call to Western Union at 1-

(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union

operator should be given Datagram Identification Number N1023 and the

following message addressed to Charles L. Miller: petitioner's name and

telephone number, date petition was mailed, plant name, and publication

date and page number of this Federal Register notice. A copy of the

petition should also be sent to the Office of the General Counsel, U.S.

Nuclear Regulatory Commission, Washington, DC 20555, and to Mark J.

Wetterhahn, Esquire, Winston and Strawn, 1400 L Street, NW.,

Washington, DC 20005-3502, attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for hearing will not

be entertained absent a determination by the Commission, the presiding

officer or the presiding Atomic Safety and Licensing Board that the

petition and/or request should be granted based upon a balancing of the

factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d)

The Commission hereby provides notice that this is a proceeding on

an application for a license amendment falling within the scope of

section 134 of the Nuclear Waste Policy Act of 1982 (NWPA), 42 U.S.C.

10154. Under section 134 of the NWPA, the Commission, at the request of

any party to the proceeding, must use hybrid hearing procedures with

respect to ``any matter which the Commission determines to be in

controversy among the parties.'' The hybrid procedures in section 134

provide for oral argument on matters in controversy, preceded by

discovery under the Commission's rules, and the designation, following

argument, of only those factual issues that involve a genuine and

substantial dispute, together with any remaining questions of law, to

be resolved in an adjudicatory hearing. Actual adjudicatory hearings

are to be held on only those issues found to meet the criteria of

section 134 and set for hearing after oral argument.

The Commission's rules implementing section 134 of the NWPA are

found in 10 CFR part 2, subpart K, ``Hybrid Hearing Procedures for

Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear

Power Reactors'' (published at 50 FR 41662, October 15, 1985) to 10 CFR

2.1101 et seq. Under those rules, any party to the proceeding may

invoke the hybrid hearing procedures by filing with the presiding

officer a written request for oral argument under 10 CFR 2.1109. To be

timely, the request must be filed within 10 days of an order granting a

request for hearing or petition to intervene. (As outlined above, the

Commission's rules in 10 CFR part 2, subpart G, and 2.714 in

particular, continue to govern the filing of requests for a hearing or

petitions to intervene, as well as the admission of contentions.) The

presiding officer shall grant a timely request for oral argument. The

presiding officer shall grant an untimely request for oral argument

only upon showing of good cause by the requesting party for the failure

to file on time and after providing the other parties an opportunity to

respond to the untimely request. If the presiding officer grants a

request for oral argument, any hearing held on the application shall be

conducted in accordance with hybrid hearing procedures. In essence,

those procedures limit the time available for discovery and require

that an oral argument be held to determine whether any contentions must

be resolved in adjudicatory hearing. If no party to the proceedings

requests oral argument, or if all untimely requests for oral argument

are denied, then the usual procedures in 10 CFR part 2, subpart G,

apply.

For further details with respect to this action, see the

application for amendment dated April 28, 1993, and revisions to this

submittal dated August 12, 1993, November 17, 1993 and February 2,

1994, which are available for public inspection at the Commission's

Public Document Room, the Gelman Building, 2120 L Street NW.,

Washington, DC 20555 and at the local public document room located at

the Salem Free Public Library, 112 West Broadway, Salem, New Jersey

08079.

Dated at Rockville, Maryland, this 25th day of February 1994.

For the Nuclear Regulatory Commission.

James C. Stone,

Project Manager, Project Directorate I-2, Division of Reactor

Projects--I/II, Office of Nuclear Reactor Regulation.

[FR Doc. 94-5026 Filed 3-3-94; 8:45 am]

BILLING CODE 7590-01-M