Applications and Amendments to Facility Operating LicensesInvolving No Significant Hazards Considerations

Federal RegisterOct 26, 1994

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NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating

LicensesInvolving No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory

Commission (the Commission or NRC staff) is publishing this regular

biweekly notice. Public Law 97-415 revised section 189 of the Atomic

Energy Act of 1954, as amended (the Act), to require the Commission to

publish notice of any amendments issued, or proposed to be issued,

under a new provision of section 189 of the Act. This provision grants

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from September 30, through October 14, 1994. The

last biweekly notice was published on October 12, 1994 (59 FR 51616).

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve 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. The basis

for this proposed determination for each amendment request is shown

below.

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

before action is taken. 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 6D22, Two White Flint North, 11545 Rockville Pike,

Rockville, 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 a hearing and petitions for leave to

intervene is discussed below.

By November 25, 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 for the particular facility involved. 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 a 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 a 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 the 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 (Project Director): 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 the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for a hearing will

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is 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 for the particular facility involved.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert

County, Maryland

Date of amendments request: September 23, 1994

Description of amendments request: The proposed amendments would

revise the Unit 2 Shutdown AC Sources Technical Specifications (TSs) to

allow a one-time extension from 7 to 14 days of the allowed outage time

(AOT) for the dedicated Class 1E emergency power source during the

upcoming Unit 2 1995 Refueling Outage (RFO-10). The proposed amendments

would also revise the Unit 1 Control Room Emergency Ventilation System

(CREVS) TSs to provide a one-time extension from 7 to 30 days of the

AOT for one train of the CREVS to be inoperable. As noted, these

extensions will be needed during the upcoming 1995 Unit 2 RFO-10 to

support the modifications scheduled for the onsite electrical

distribution system in response to the Station Blackout (SBO) Rule, 10

CFR 50.63, and the upgrade of No. 21 Emergency Diesel Generator

(EDG).The specific changes requested are:

Unit 2 TSs 3.8.1.2 and 3.8.2.2 will include a footnote indicating

that the AOT for aligning an operable emergency diesel generator (EDG)

to provide power to the emergency busses within 14 days during the Unit

2 RFO-10.

Unit 1 TS 3.7.6.1 will be modified to indicate that during the No.

21 EDG upgrade, the time to restore the No. 21 filter train of the air

conditioning unit to operable status may be extended to 30 days (for

loss of emergency power only) if: 1) A temporary diesel generator is

demonstrated to be available by starting it at least once per 7 days

and 2) if action 1 is not met, restore compliance with the action

within 7 days or be at least in hot standby within the next 6 hours and

in cold shutdown within the following 30 hours.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of issue of no significant hazards consideration

for each of the proposed changes, which is presented below:

In relation to the requested changes to the Unit 2 TSs:

1. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

Requiring one Class 1E Emergency Diesel Generator (EDG) to be

available for a shutdown unit ensures that AC power will be

available for a loss of offsite power event, a boron dilution event,

or a fuel handling incident. There is a very low probability that a

loss of offsite power will occur due to severe weather or

inadvertent damage to the switchyard during the 14-day period that

the temporary splice box is being installed and No. 12 EDG is out-

of-service. The Calvert Cliffs offsite power supply is highly

redundant and has significant capability in withstanding severe

weather events, such as tornadoes. In addition, Calvert Cliffs

Emergency Response Plan Implementation Procedures requires that

certain actions be taken, up to and including shutdown of both

units, on the approach of a severe storm, such as a hurricane. The

probability of a loss of offsite power is maintained low by

prohibiting planned maintenance on two of the three 500 kV

transmission lines and associated relaying and devices within the

switchyard. Availability of the required offsite power sources will

be verified once per shift. In addition to the offsite power

sources, a temporary diesel generator will also be installed to

provide a backup onsite power source with the capacity to support

the safety-related loads of the shutdown unit.

The boron dilution event and the fuel handling incident are the

only two accidents that are explicitly analyzed in the Updated Final

Safety Analysis Report for a shutdown unit. The potential accident

precursors such as core alterations, positive reactivity insertions,

movement of irradiated fuel and movement of heavy loads over

irradiated fuel, will be prohibited while No. 12 EDG is out-of-

service for the temporary splice box installation. Therefore the

probability of a boron dilution event or fuel handling incident is

decreased during the operations allowed by this change. The

requirement to maintain containment penetration closure ensures that

the consequences of an accident would not be significantly

increased.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2. Would not create the possibility of a new or different type

of accident from any accident previously evaluated.

A temporary diesel generator is being installed onto a 4 kV bus

of the shutdown unit while the dedicated EDG for this unit is

transferred to the operating unit for up to 14 days. This is an

extension of the same configuration allowed by Action Statements

3.8.1.2.b and 3.8.2.2.b with additional provision taken for the

Control Room Emergency Ventilation System (CREVS). The EDGs will be

aligned so that each train of the CREVS will have an emergency power

supply available. The proposed change has been evaluated and it has

been determined that it does not impair any existing safety-related

equipment needed to maintain the unit in a safe shutdown condition,

and does not create any new accident initiators. The operation of

the temporary diesel generator is familiar to the operators and is

not significantly different from typical operator activities.

Therefore, the proposed change does not create the possibility

of a new or different type of accident from any accident previously

evaluated.

3. Would not involve a significant reduction in a margin of

safety.

The safety function provided by the AC electrical power sources

and associated distribution systems for a shutdown unit is to ensure

that the unit can be maintained in a safe shutdown condition, and

there is sufficient instrumentation and control capability available

for monitoring and maintaining the unit status. The proposed change

would allow the shutdown unit to be without a dedicated Class 1E

emergency power source for up to 14 days. This is an extension of

the outage time of seven days allowed by the Technical

Specifications for performing maintenance and inspections on No. 12

EDG. This proposed change will have no impact on the offsite power

sources.

Several compensatory measures will be taken during this period

to ensure that a power source will be available for the shutdown

unit. These measures include requiring that two offsite power

sources are available, and a temporary diesel generator will be

installed capable of supplying the loads necessary to maintain the

unit in a safe condition. In addition, Technical Specifications

require several compensatory measures to reduce the potential for a

fuel handling incident and a boron dilution event. These measures

include prohibiting positive reactivity changes, suspending core

alterations, movement of irradiated fuel, and the movement of heavy

loads over irradiated fuel. Establishing containment penetration

closure further ensures that adequate margin of safety is

maintained. In addition, reduced inventory conditions of the Reactor

Coolant System will be prohibited during the 14-day period.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

In relation to the requested changes in the Unit 1 TSs:

1. Would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The Control Room Emergency Ventilation System (CREVS) is

designed so that the Control Room can be occupied under all plant

conditions. The CREVS is required to maintain the Control Room

temperature and to filter the Control Room air in the event of a

radioactive release. When No. 21 Emergency Diesel Generator (EDG) is

being upgraded, No. 12 CREVS will be without a Class 1E emergency

power source. The CREVS is not an initiator in any previously

evaluated accidents. Therefore, the proposed change does not involve

an increase in the probability of an accident previously evaluated.

The CREVS is required to maintain the Control Room habitable

following a radioactive release from a loss of coolant accident, a

main steam break, or a steam generator tube rupture. There is a very

low probability of an event occurring requiring Control Room

isolation during the 30-day period that it will take to upgrade No.

21 EDG. Requiring that the CREVS have both a normal power source and

an emergency power source available ensures that one train of the

system will be available so that the Control Room can be occupied

under these conditions. The probability of a loss of offsite power

is very low due to the highly redundant design of the offsite power

supply. Planned maintenance on three of the offsite power supplies

and associated relaying and devices within the switchyard will be

prohibited during the upgrade period to maintain the low probability

of a loss of offsite power event. Number 12 CREVS train will

continue to have its normal power source for all but approximately

four days when the bus will be de-energized to allow bus work that

is necessary to the tie-in of the Alternate AC diesel generator.

Number 11 CREVS will have both its normal and emergency power supply

available and this train is capable of maintaining the Control Room

habitable. In addition, a temporary diesel generator will be

installed to provide assurance that an emergency power source will

be available to No. 12 CREVS. The compensatory measures that will be

taken during this period will ensure that the proposed change does

not involve a significant increase in the consequences of an

accident previously evaluated.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2. Would not create the possibility of a new or different type

of accident from any accident previously evaluated.

The CREVS is not being modified by this proposed change. The

system will continue to operate in the same manner. Number 21 EDG

will operate in a similar manner after the upgrade and will be able

to support unit operation after all the testing is completed. The

installation of the temporary diesel generator during the upgrade

period has been evaluated to ensure that it does not create any new

accident initiators.

Therefore, the proposed change does not create the possibility

of a new or different type of accident from any accident previously

evaluated.

3. Would not involve a significant reduction in a margin of

safety.

The operability of the CREVS during Modes 1 through 4 ensures

that the Control Room will remain habitable under all plant

conditions. The proposed change does not affect the function of the

CREVS. The proposed change will allow one train of the CREVS to be

without a Class 1E emergency power supply for up to 30 days. This

train will have the normal power supply available for all but

approximately four days to allow necessary bus work. The other train

of the CREVS will have both its normal and emergency power supplies

during this period. Compensatory measures that will be taken include

prohibiting planned maintenance on the required offsite power

sources and installing a temporary diesel generator of sufficient

capacity as a backup to the affected train. These measures will

maintain the current margin of safety. The upgrade to the existing

EDGs will provide additional margin for the electrical loading of 4

kV safety-related busses. The completion of the No. 21 EDG upgrade

will improve the margin of safety for the onsite electrical

distribution system.

Therefore, the proposed change does 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 50.92(c) are

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

amendments request involves no significant hazards consideration.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,

Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Project Director: Ledyard B. Marsh

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of amendment request: September 6, 1994

Description of amendment request: The proposed amendment changes

the Pilgrim Nuclear Power Station Technical Specifications Sections

3.7.B.1.a, 3.7.B.1.c, 3.7.B.1.e, 3.7.B.2.a, and 3.7.B.2.c. The proposed

changes also add new sections 3.7.B.1.f and 3.7.B.2.e. These sections

require both trains of the Standby Gas Treatment (SGTS) and Control

Room High Efficiency Air Filtration (CRHEAF) System to be operable for

the initiation of fuel movement and during fuel handling operations

involving irradiated fuel.

Basis for proposed no significant hazards consideration

determination: 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. The operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Technical Specifications 3.7.B.1 and 3.7.B.2.e restrict the

movement of irradiated fuel when only one train of SGTS or one train

of CRHEAF are operable. Irradiated fuel movement may not begin and

may only continue for seven days when the Limiting Condition of

Operation is entered.

Removing these restrictions during refueling operations does not

involve a significant increase in the probability or consequences of

an accident previously evaluated because compensatory measures will

be in place.

When sections 3.7.B.1.f and 3.7.B.2.e are invoked fuel movement

will not commence until 5 days following plant shutdown and reactor

vessel will be flooded-up to elevation 114''. The 5 day period

provides decay-time before irradiated fuel movement begins.

Flooding-up elevation 114'' provides an enlarged inventory reducing

the possibility of a loss-of-coolant event exposing fuel such that

radioactive gasses are produced, an event SGTS and CRHEAF are

designed to mitigate.

Other compensatory measures include requiring the SBO [station

blackout] diesel or the shutdown transformer to be operable prior to

and during the fuel movement. This adds defense-in-depth by making

available another power supply to the in-service safety-related bus.

Also, the substitution of a non-safety power supply to the SGTS and

CRHEAF ``inoperable'' systems while their safety-grade bus is out-

of-service for maintenance will provide offsite power to the

``inoperable'' train. While this electrical supply is not safety-

grade, it is reliable and capable of powering the SGTS and CRHEAF

systems. The components of the ``inoperable'' trains will be

available with power from an alternate power source. The

compensatory connection to the non-safety grade bus gives added

confidence these trains can perform the design function although

they are not ``operable'' as defined by Technical Specifications.

Operating Pilgrim in accordance with this proposed change does

not involve a significant increase in the probability or consequence

of an accident previously analyzed because compensatory measures

will be in force to: restrict the commencement of irradiated fuel

handling or new fuel handling over the spent fuel or core until 5

days following reactor shutdown; provide a reliable source of power

to the ``inoperable'' SGTS and CRHEAF systems; provide an enlarged

coolant inventory to protect irradiated fuel from the effects of an

inadvertent draindown of the vessel; and provide an additional

source of emergency power to the active SGTS and CRHEAF systems by

ensuring the operability of the SBO diesel generator or the Shutdown

Transformer.

2. The operation of Pilgrim Station in accordance with the

proposed amendment will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Planned maintenance activities require removing a safety-related

bus and emergency diesel generator powering a train of SGTS and

CRHEAF from service. The redundant trains are not affected. The

affected trains of SGTS and CRHEAF will be connected to a non-safety

bus, allowing them to operate but not allowing them to be considered

operable under the purview of Technical Specifications. The proposed

change allows refueling activities to commence with one train of

SGTS and CRHEAF fully operable and the other train available but not

generator. Compensatory measures will be in effect during refueling

activities involving this configuration. The proposed changes do not

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

the fuel-drop accident previously analyzed. Therefore, operating

Pilgrim in accordance with this change will not create the

possibility of a new or different kind of accident from any accident

previously analyzed.

3. The operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant reduction in the

margin of safety.

SGTS and CRHEAF contribute to the margin of safety during fuel

handling by mitigating the consequences of a fuel-handling event.

Allowing an exception to the requirement of both trains of SGTS and

CRHEAF operable prior to or during fuel movement activities does not

involve a significant reduction in the margin of safety because the

first line of defense, the other SGTS and CRHEAF trains, will be

operable. The redundant trains will also be powered and operable in

all ways except the ``operable'' concept required by Technical

Specification.

Hence, the actual condition of the equipment allows it to meet

its design function except under the strict Technical Specification

interpretation of operable, and the described compensatory measures

that will be in effect when the exception is employed, constrain the

potential impact on the margin of safety caused by using the

exception; therefore, operating Pilgrim in accordance with this

proposed Technical Specification request does not involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,

800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.

NRC Project Director: Walter R. Butler

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of amendment request: September 6, 1994

Description of amendment request: The proposed amendment would

reduce the Reactor Pressure Setpoint at which the shutdown cooling

system automatically isolates. This setpoint also isolates the low

pressure coolant injection valves when the shutdown cooling system is

in operation.

Basis for proposed no significant hazards consideration

determination: 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. The operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Technical Specification Table 3.2.A lists the instrumentation

that initiates primary containment isolation and also lists the trip

level setting (setpoints) for that instrumentation. The setpoint for

reactor high pressure is presently [less than or equal to] 110 psig

which was selected to provide protection for the RHR [residual heat

removal] low pressure suction piping against possible

overpressurization. This signal initiates a group 3 containment

isolation by closing the shutdown cooling isolation valves and the

Low Pressure Coolant Injection (LPCI) valves. To provide an optimal

solution to address Generic Letter 89-10, the motor-operated valves

which effect the isolation of the RHR suction piping (MO1001-47 and

MO1001-50) are being modified based on a lower differential pressure

in the design calculations. The setpoint is being reduced to ensure

plant operation is maintained in accordance with the new design and

to continue to provide the protection necessary against

overpressurization. This does not involve an increase in the

probability or consequences of an accident previously analyzed

because reducing the setpoint to less than what the technical

specifications currently requires is a change in the conservative

direction relative to protection of the piping. The LPCI injection

valves are designed for higher pressures and the proposed setpoint

change does not involve an increase in the probability or

consequences of an accident previously evaluated.

Technical Specification Table 3.2.B lists instrumentation that

initiates or controls the core and containment cooling systems and

also lists the trip level settings (setpoints) for that

instrumentation. The setpoint for reactor low pressure [less than or

equal to] 110 psig, is a permissive for the group 3 isolation of the

RHR inboard injection valves. Reducing the setpoint to [less than or

equal to] 76 psig is consistent with the design of the other group 3

isolation valves that receive the same signal and accomplishes the

isolation of the shutdown cooling system when there is a system

breach. Thus, revising this setpoint does not increase the

probability or consequences of an accident previously evaluated.

2. The operation of Pilgrim Station in accordance with the

proposed amendment will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed setpoint change supports modifications made to the

shutdown cooling isolation valves to provide additional margin to

address Generic Letter 89-10 concerns. Reducing the setpoint for

this function continues to provide protection of the RHR suction

piping and ensures closure of the isolation valves. Therefore,

revising the reactor high pressure setpoint to [less than or equal

to] 76 psig for instrumentation that initiates primary containment

isolation (Table 3.2.A) does not create the possibility of a new or

different kind of accident previously evaluated. Similarly, the

revision of the reactor low pressure setpoint to [less than or equal

to] 76 psig for instrumentation that initiates or controls the core

and containment cooling systems does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. The operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant reduction in a

margin of safety.

The purpose of the setpoint for reactor pressure in Table 3.2.A

and 3.2.B is to provide protection for the RHR suction piping and

ensure proper isolation for unlikely piping breaches. Changing the

setpoint to a lower value is consistent with modifications being

made to the shutdown cooling isolation valves. The margin of safety

for this setpoint was established to protect the RHR suction piping

from overpressurization and to ensure that primary containment

integrity could be established by the isolation valves on a Group 3

isolation. A margin of safety for protecting the RHR suction piping

exists due to the difference between the design pressure of the

piping and the setpoint specified in the technical specifications.

Reducing the setpoint increases the difference between the design

pressure of the piping and the setpoint hence, this margin of safety

is increased. The margin of safety established for primary

containment isolation valves is maintained by specifying a setpoint

which corresponds to the closing differential pressure of the valves

under postulated accident conditions. The setpoint change does not

reduce the design margins established to ensure the valves perform

their design isolation function when required. The low pressure

coolant injection valves that receive this same signal are designed

for higher pressures than the current setpoint of [less than or

equal to] 110 psig and, therefore, a lower setpoint increases the

margin of safety. Thus, the proposed amendment does 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 50.92(c) are

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,

800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.

NRC Project Director: Walter R. Butler

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of amendment request: September 6, 1994

Description of amendment request: The proposed amendment would

remove Technical Specification section 4.5.H.4, a section which

requires the testing and calibration of pressure switches in certain

emergency core cooling system (ECCS) lines.

Basis for proposed no significant hazards consideration

determination: 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. The Operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

[***].

The discharge piping for ECC systems is maintained filled to

prevent water hammer during automatic pump starts. Monthly venting

is the primary means of ensuring filled discharge piping. The

pressure switches are an adjunct to such venting. Hence, piping in

the Core Spray System, the Low Pressure Coolant Injection System

(LPCI), the High Pressure Coolant Injection (HPCI) system, and the

Reactor Core Isolation Coolant (RCIC) system are all equipped with

pressure switches that detect pressure decay in the discharge piping

of these systems.

This proposed change does not change Pilgrim's configuration or

equipment. The switches perform a surveillance function and do not

provide a signal needed to prevent or mitigate an accident. The

switches will continue to perform their surveillance function and

their surveillance and calibration will be performed in accordance

with Pilgrim procedures. Removal of section 4.5.H.4 eliminates the

possibility of inoperable switches forcing the shutdown of Pilgrim

or the alternative of declaring an operable safety system inoperable

because of its association with these switches.

Technical Specifications will continue to require venting the

discharge piping high point when the systems are configured such

that water hammer can occur. (sections 4.5.H.1, 4.5.H.2 and

4.5.H.3). Thus, the application of this proposed change does not

reduce the Technical Specifications intent of reducing the

likelihood of discharge piping water hammer. Therefore, operating

Pilgrim Station in accordance with the proposed amendment will not

involve a significant increase in the probability or consequences of

an accident previously analyzed.

2. The operation of Pilgrim Station in accordance with the

proposed amendment will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Section 4.5.H's purpose is to maintain the ECCS discharge piping

filled to prevent water hammer. The purpose of the pressure switches

is to detect voids in ECCS discharge piping to prevent the

possibility of damage due to water hammer. These switches are not

safety-related, have no automatic functions, and are not relied on

to prevent or mitigate an accident. Instead, they enhance the

existing discharge pipe venting surveillance requirements by

detecting void formation in discharge pipe.

The switches will continue to perform their surveillance

function through Pilgrim procedures. Venting will continue to be

required by Technical Specifications. Therefore, operating Pilgrim

in accordance with this proposed change does not create the

possibility of a new or different kind of accident from any accident

previously evaluated because the proposed change does not impair the

detection of conditions necessary to produce a water hammer in the

discharge piping.

3. The operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant reduction in a

margin of safety.

The discharge piping pressure switches are surveillance

instruments and act as a secondary means of protecting the discharge

piping from conditions that can produce water hammer. They are not

relied on to prevent or mitigate accidents. Hence, these switches do

not significantly impact safety because they are not the primary

means of preventing discharge piping water hammer. Therefore,

removing the pressure switches from Technical Specifications

potentially contributes to plant availability but does not involve a

significant reduction in a margin of safety because the primary

method of detection (venting) remains and the switches will continue

to be subject to procedural controls.

This proposed change has been reviewed and recommended for

approval by the Operations Review Committee and reviewed by the

Nuclear Safety Review and Audit Committee.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,

800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.

NRC Project Director: Walter R. Butler

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of amendment request: September 6, 1994

Description of amendment request: The proposed amendment would

relocate the alarms for the drywell to suppression chamber vacuum

breakers to a different annunciator panel.

Basis for proposed no significant hazards consideration

determination: 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. The operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant increase in the

probability or consequences of an accident previously identified.

The proposed change relocates annunciators in the control room

but does not change their designed function or setpoint.

The Annunciator System is non-safety related and performs no

direct safety function. No accident initiators are being affected by

this proposed change. Accident mitigating systems remain operable,

and accident scenarios are unaffected.

2. The operation of Pilgrim Station in accordance with the

proposed amendment will not create the possibility of a new or

different kind of accident from any accident previously analyzed.

Relocating the Drywell to Suppression Chamber annunciator from

one control room panel to another does not create the possibility of

a new or different kind of accident. This modification does not

modify the setpoints or functions of the annunciators. Hence, it is

administrative and proposed to allow relocation which is currently

constrained by the current Technical Specifications level of detail.

3. The operation of Pilgrim Station in accordance with the

proposed amendment will not involve a significant reduction in the

margin of safety.

The equipment being relocated is non-safety related and its

relocation does not impact the margin of safety. This relocation is

proposed to enhance the operator's ability to identify and analyze

abnormal events.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,

800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.

NRC Project Director: Walter R. Butler

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois;Docket Nos.

STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,

Will County, Illinois

Date of amendment request: June 13, 1994, as supplemented on

October 7, 1994.

Description of amendment request: The proposed amendment would make

several changes to the Administrative Controls in Section 6 of

Technical Specifications (TS) for Byron and Braidwood stations. The

proposed changes include: (1) a change to the submittal frequency of

the Radiological Effluent Release Report, (2) a revision to the Shift

Technical Advisor description, (3) clarification of the Shift

Engineer's responsibilities, and (4) editorial changes. The references

to the Semiannual Radiological Effluent Release Report are also revised

in other sections of the TS. The proposed change in the October 7,

1994, submittal revised TS 6.3.1 to include generic descriptions of

personnel who fulfill the responsibilities of a radiation protection

manager. This supplements the information that was published in the

Federal Register on August 3, 1994 (59 FR 39581).

Basis for proposed no significant hazards consideration

determination: 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:

A. The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed changes to Section 6 of Technical Specifications do

not affect any accident initiators or precursors and do not change

or alter the design assumptions for the systems or components used

to mitigate the consequences of an accident.

The proposed changes are administrative in nature and provide

clarification. These changes provide consistency with station

procedures, programs, the Code of Federal Regulations, other

Technical Specifications, and Standard Technical Specifications.

These changes do not impact any accident previously evaluated in the

Updated Final Safety Analysis Report.

B. The proposed changes do not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed changes do not affect the design or operation of

any system, structure, or component in the plant. There are no

changes to parameters governing plant operation; no new or different

type of equipment will be installed. The proposed changes are

considered to be administrative changes. All responsibilities

described in Technical Specifications for management activities will

continue to be performed by qualified individuals.

C. The proposed changes do not involve a significant reduction

in a margin of safety.

The proposed changes do not affect the margin of safety for any

Technical Specification. The initial conditions and methodologies

used in the accident analyses remain unchanged, therefore, accident

analysis results are not impacted.

The proposed changes are administrative in nature and have no

impact on the margin of safety of any Technical Specification. They

do not affect any plant safety parameters or setpoints. The

descriptions for the Shift Technical Advisor and Shift Engineer are

clarified, however, include no reduction to their responsibilities.

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.

Local Public Document Room location: For Byron, the Byron Public

Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for

Braidwood, the Wilmington Township Public Library, 201 S. Kankakee

Street, Wilmington, Illinois 60481.

Attorney for licensee: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60690

NRC Project Director: Robert A. Capra

Georgia Power Company, Oglethorpe Power Corporation, Municipal

Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.

50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,

Burke County, Georgia

Date of amendment request: September 13, 1994

Description of amendment request: The amendments replace

Containment Systems technical specification (TS) 3.6.2.2, ``Spray

Additive System'' with a new Emergency Core Cooling Systems TS 3.5.5,

``ECCS Recirculation Fluid pH Control System.''

Basis for proposed no significant hazards consideration

determination: 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. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The proposed change involves replacement of concentrated

NaOH injected via the containment spray system with trisodium

phosphate (TSP) stored in the containment and dissolved in the sump

recirculation solution to maintain acceptable post accident spray/

recirculation solution chemistry. Deletion of the concentrated NaOH

will eliminate a personnel hazard. The pH control system functions

in response to an accident and does not involve or have any effect

on any initiating event for any accident previously evaluated.

Operation under the proposed amendment will continue to ensure that

iodine potentially released post-LOCA is retained in the sump

solution, and resultant offsite and control room thyroid doses are

within the limits of 10 CFR 100 and 10 CFR 50, Appendix A, General

Design Criterion 19, respectively.

2. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated. The deleted equipment is isolated from the remaining

equipment by blind flanges, locked closed valves, cut and capped

piping, determinated and/or spared cables; and interfaces are

analyzed to ensure the remaining required equipment meets applicable

original design requirements. The new equipment (TSP and baskets) is

a passive pH control system and is supported and analyzed to ensure

there are no adverse interfaces (e.g. pipe break, jet impingement,

seismic) with existing equipment, systems, or structures.

3. The proposed change does not involve a significant reduction

in a margin of safety. The slight change in recirculation solution

pH maintains adequate protection against chloride induced stress

corrosion cracking of austenitic stainless steel and maintains the

capability of the solution to retain iodine. It results in an

insignificant increase in the post-accident rate of hydrogen

generation, which remains well within the existing capacity of the

hydrogen recombiners. The increased mass in the containment will

have no significant impact on post-accident flood levels,

recirculation solution boron concentration, or peak clad

temperatures. No other operating parameters for systems, structures,

or components assumed to operate in the safety analysis are changed.

The offsite and control room doses meet the limits of 10 CFR 100 and

GDC 19 respectively. Because the trisodium phosphate is nonvolatile

and the baskets are protected with solid covers and are located

slightly above the floor in the containment where access is strictly

controlled, a surveillance interval of once per refueling outage

provides assurance that the TSP will be available when required.

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.

Local Public Document Room location: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia 30830.

Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,

Georgia 30308

NRC Project Director: Herbert N. Berkow

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek

Nuclear Generating Station, Ocean County, New Jersey

Date of amendment request: September 26, 1994

Description of amendment request: The proposed license amendment

would revise the ``Plan for the Long Range Planning Program'' by

changing the semi-annual reporting period to annual, and to reflect

refined evaluation criteria and assessment methodology; and, to

incorporate the necessary changes to the license condition wording.

Basis for proposed no significant hazards consideration

determination: 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. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or consequences of an accident previously

evaluated. The proposed revision to the Facility Operating License

does not affect the safety analysis and does not involve any

physical changes to the plant, nor any changes in the format or

restraints on plant operations, and only contemplates a change to

the Plan for the Long Range Planning Program currently approved by

the NRC in license condition 2.C.(6). Therefore, this change will

not increase the probability of previously analyzed accidents

because it involves no direct plant modification or change in

operation, and hence, it is also unrelated to the possibility of

increasing the consequences of previously analyzed accidents.

2. Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different

kind of accident from any previously evaluated. The proposed

revision to the Facility Operating License does not affect the

safety analysis and does not involve any physical changes to the

plant, nor any changes in the format or restraints on plant

operations, and only contemplates a change to the Plan for the Long

Range Planning Program currently approved by the NRC in license

condition 2.C.(6). Therefore, this change has no effect on the

possibility of creating a new or different kind of accident from any

previously evaluated.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety. The proposed revision to the Facility Operating License does

not involve any physical changes to the plant, nor any changes in

the format or restraints on plant operations, and only contemplates

a change to the Plan for the Long Range Planning Program currently

approved by the NRC, in license condition 2.C.(6). Therefore, the

overall margin of safety for the plant is maintained.

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.

Local Public Document Room location: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, NJ 08753

Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,

Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Project Director: Phillip F. McKee

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile

Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: September 26, 1994

Description of amendment request: The proposed license amendment

would revise the ``Plan for the Long Range Planning Program'' by

changing the semi-annual reporting period to annual, and to reflect

refined evaluation criteria and assessment methodology; and, to

incorporate the necessary changes to the license condition wording.

Basis for proposed no significant hazards consideration

determination: 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. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the

probability of occurrence or consequences of an accident previously

evaluated. The proposed revision to the Facility Operating License

does not affect the safety analysis and does not involve any

physical changes to the plant, nor any changes in the format or

restraints on plant operations, and only contemplates a change to

the Plan for the Long Range Planning Program currently approved by

the NRC in license condition 2.C.(9). Therefore, this change will