Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No Significant Hazards Considerations

Federal RegisterMar 11, 1998

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NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving 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 February 13, 1998, through February 27,

1998. The last biweekly notice was published on February 25, 1998 (63

FR 9589).

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 Chief, Rules and

Directives Branch, Division of Administration Services, Office of

Administration, U.S. Nuclear Regulatory Commission, Washington, DC

20555-0001, 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

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

The filing of requests for a hearing and petitions for leave to

intervene is discussed below.

By April 10, 1998, 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 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 20555-0001, Attention: Rulemakings and

Adjudications Staff, or may be delivered to the Commission's Public

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

by the above date. A copy of the petition should also be sent to the

Office of the General Counsel, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, 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, and at the local public document room for

the particular facility involved.

Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,

Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

Date of amendment request: January 14, 1998, which superseded the

September 3, 1997, submittal.

Description of amendment request: The proposed amendments would

revise the Technical Specifications to reduce the allowable Unit 1

Reactor Coolant System Dose Equivalent Iodine-131 from 0.35

microCuries/gram to 0.05 microCuries/gram thru the end of Unit 1, Cycle

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.

Generic Letter 95-05, ``Voltage-Based Repair Criteria For

Westinghouse Steam Generator Tubes Affected By Outside Diameter

Stress Corrosion Cracking,'' allows lowering of the RCS [Reactor

Coolant System] DE-131 [Dose Equivalent Iodine-131] activity as a

means for accepting higher projected leak rates if justification for

equivalent I-131 below 0.35 microCuries/gram is provided. Four

methods for determining the impact of a release of activity to the

public were reviewed to provide this justification. These four

methods are as follows:

Method 1: NRC NUREG 0800, Standard Review Plan (SRP) Methodology

[[Page 11915]]

Method 2: Methodology described in a report by J.P. Adams and C.L.

Atwood, ``The Iodine Spike Release Rate During a Steam Generator

Tube Rupture,'' Nuclear Technology, Vol. 94, p. 361 (1991) using

Braidwood Station reactor trip data.

Method 3: Methodology described in a report by J.P. Adams and C.L.

Atwood, ``The Iodine Spike Release Rate During a Steam Generator

Tube Rupture,'' Nuclear Technology, Vol. 94, p. 361 (1991) using

normalized industry reactor trip data.

Method 4: Methodology described in a draft EPRI Report TR-103680,

Revision 1, November 1995, ``Empirical Study of Iodine Spiking in

PWR Plants''.

The effect of reducing the RCS DE I-131 activity limit on the

amount of activity released to the environment remains unchanged

when the maximum site allowable primary-to-secondary leak rate is

proportionately increased and the iodine release rate spike factor

is assumed to be 500 in accordance with the SRP. With an RCS DE I-

131 activity limit of 1.0 microCuries/gram, the maximum site

allowable leakage limit was calculated, in accordance with the NRC

SRP methodology, to be 6.64 gpm at room temperature and pressure.

ComEd has evaluated the reduction of the RCS DE I-131 activity to

0.05 microCuries/gram along with the increase of the allowable

leakage to 132.8 gpm at room temperature and pressure and has

concluded:

--assuming a spike factor of 500, the maximum activity released is

not changed, and

--the offsite dose, including the iodine spiking factor, will be

less than the 10 CFR 100 limits.

Based on the NRC SRP methodology for dose assessments and

assuming the iodine spike factor of 500 is applicable at the new

0.05 microCuries/gram RCS DE I-131 activity limit, the Control Room

dose, the Low Population Zone dose, and the dose at the Exclusion

Area Boundary continue to satisfy the appropriately small fraction

of the 10 CFR 100 dose limits.

An evaluation of the Control Room dose, attributed to an MSLB

accident concurrent with steam generator primary-to-secondary

leakage at the maximum site allowable limit, was performed in

support of a license amendment request for application of a 1.0 volt

Interim Plugging Criteria. This evaluation concluded that the

activity released to the environment during an eight (8) hour time

period from an MSLB accident (812 Curies for a Pre-accident iodine

spike and 888 Curies for an accident-initiated iodine spike) is

bounded by the activity released to the environment from the Loss of

Coolant design basis accident (1290 Curies). Therefore, the Control

Room dose, due to the MSLB accident scenario, is bounded by the

existing Loss of Coolant Accident (LOCA) analysis. The maximum site

allowable primary-to-secondary leakage is limited by the offsite

dose at the Exclusion Area Boundary due to an accident-initiated

spike.

The report by J.P. Adams and C.L. Atwood, ``The Iodine Spike

Release Rate During a Steam Generator Tube Rupture,'' Nuclear

Technology, Vol. 94, p. 361 (1991), concluded that the NRC SRP

methodology, which specifies a release rate spike factor of 500 for

iodine activity from the fuel rod to the RCS, is conservative when

the RCS DE I-131 concentration is greater than 0.3 microCuries/gram.

In order to evaluate whether a release rate spike factor of 500 is

conservative below 0.3 microCuries/gram, actual operating data from

the previous reactor trips of Braidwood Units 1 and 2, with and

without fuel defects, were reviewed and analyzed using the

methodology presented in Section II.C of the Adams and Atwood report

(Method 2). The same five data screening criteria described in the

Adams and Atwood report were applied to the Braidwood data to ensure

consistency and validity when comparing the Braidwood results to the

data in the Adams and Atwood report. Of the reactor trip events at

Braidwood Units 1 and 2, seventeen (17) met the five data screening

criteria.

Seven (7) of the seventeen (17) Braidwood trips occurred during

cycles with no fuel defects. In all seven of these instances, the

calculated spike factor was much less than the spike factor of 500

assumed in the NRC SRP methodology. Braidwood Unit 1 Cycle 7 is

currently operating with no fuel defects and an RCS DE I-131

activity of approximately 3E-4 microCuries/gram. The seven previous

trips with no fuel defects had steady-state iodine values that are

reasonably close to the current operating conditions. It is

therefore reasonable to conclude that, assuming continued operation

with little to no fuel defects, the calculated spike factors from

these events would reflect an actual event for Unit 1 Cycle 7, i.e.

the spike factor will be less than 500.

Since some of the Braidwood spike factors were greater than 500

when the RCS DE I-131 activity prior to the accident was less than

0.3 microCuries/gram, ComEd examined the conservatisms in the

current release rate calculation. The primary reason for the high

spiking factors contained in the Adams and Atwood report (up to

12,000), is not because the absolute post-trip release rate is high

(factor numerator), but rather because the steady-state release rate

(factor denominator) is low. The Braidwood specific data resulted in

six (6) events with a calculated release rate spike factor greater

than 500. It is not expected based upon the Unit 1 Cycle 7 fuel

conditions that a spiking factor greater than 500 would occur. The

revised RCS DE I-131 activity limit will also ensure that the

operating cycle will not continue if significant fuel defects

develop.

In order to evaluate the Braidwood specific data against the NRC

SRP methodology, the release rate for a steady-state RCS DE I-131

activity of 1.0 microCuries/gram was calculated. Using the Braidwood

specific data, the pre-trip steady-state release rate is 27.5 Ci/hr.

Using a release rate spike factor of 500 for the accident-initiated

spike, the post-trip maximum release rate would be 13,733 Ci/hr (SRP

Methodology). The highest post-trip iodine release rate from the

Braidwood trip data, Event 15, was 1335 Ci/hr, it is important to

remember that this number is determined by conservatively increasing

the post-trip RCS DE I-131 activity by a factor of three (3), in

accordance with the Adams and Atwood report.

The purpose of this amendment request is to reduce the TS

[Technical Specification] RCS DE I-131 limit by a factor of twenty

as compared to the original TS RCS DE I-131 limit of 1.0

microCuries/gram. By decreasing the TS RCS DE I-131 activity by a

factor of twenty the maximum iodine release rate is 686.7 Ci/hr,

(13,733 Ci/hr divided by 20). Two (2) of the seventeen (17)

Braidwood data points exceed this value. Both occurred during cycles

with fuel defects. Braidwood Unit 1 is currently operating with no

fuel defects. Fifteen (15) of the 168 data points in the Adams and

Atwood report exceed 686.7 Ci/hr. For the combined database of 185

data points, of which 17 exceeded 686.7 Ci/hr, only two of these

seventeen (17) data points had a pre-trip RCS DE I-131 activity

below 0.05 microCuries/gram. The 95% confidence prediction for the

combined data sets bounded one (1) of these two (2) data points.

This data indicates that the possibility for a post-trip iodine fuel

release rate to exceed 686.7 Ci/hr, when the pre-trip RCS DE I-131

concentration is at or below 0.05 microCuries/gram, is small. The

conservatisms mentioned in the following sections will reduce the

possibility of exceeding a small fraction of the 10 CFR 100 limits

should a fuel release greater than 686.7 Ci/hr occur.

If the Braidwood data were plotted with the Adams and Atwood

data, the conclusions of the Adams and Atwood report would not be

compromised. Where the Braidwood data contains spike factors greater

than 500, the RCS DE I-131 concentrations are below 0.05

microCuries/gram. Since the Braidwood data includes very few data

points near 0.05 microCuries/gram (the requested new TS limit), it

is appropriate to use the Braidwood database combined with the Adams

and Atwood database near 0.05 microCuries/gram to determine if a

spike factor of 500 is appropriate. The combined databases contain

seventy-nine (79) data points with a Pre-Trip RCS DE I-131 activity

between 0.01 microCuries/gram and 0.10 microCuries/gram. Sixty-two

(62) of these seventy-nine (79) data points (78%) have spike factors

less than 500. Using the entire Braidwood database combined with the

Adams and Atwood database, 141 of the 185 data points (76%) have an

iodine spike factor less than 500. Therefore, it is reasonable to

assume that a spike factor of 500 would not be exceeded for a

majority of the events if an MSLB accident were to occur while the

RCS DE I-131 activity is at or below 0.05 microCuries/gram. The

highest spike factor seen in the Adams and Atwood report near a Pre-

Trip RCS DE I-131 activity of 0.05 microCuries/gram was 773 (at 0.05

microCuries/gram). The corresponding release rate for this event was

368 Ci/hr which is less than the calculated Braidwood maximum

release rate of 686.7 Ci/hr.

The predominant factors in calculating the offsite dose are the

post-trip iodine release rate from the fuel and the flowrate at

which the activity is being released to the environment, not whether

the spike factor is greater than or less than 500. The post-trip DE

I-131 release rate will determine the level of activity in the RCS

that will be released. The flowrate will determine at what rate this

[[Page 11916]]

activity is released to the environment. Method 3, which used an

approach in the Adams and Atwood report, concluded that, at a 95%

confidence of a 85 percentile, the post-trip iodine release rate was

bounded by 0.608 Ci/hr-MWe. For Braidwood Station, which has a MWe

rating of 1175, the post-trip iodine release rate, at a 95%

confidence of a 85 percentile, should not exceed 714 Ci/hr. Two (2)

of the seventeen (17) reactor trips from Braidwood exceeded 714 Ci/

hr. These two (2) reactor trips had post-trip iodine release rates

of 1335 Ci/hr (spike factor of 3471) and 802 Ci/hr (spike factor of

1483). Both occurred during cycles with fuel defects. Braidwood Unit

1 is currently operating with no fuel defects.

In the fourth method, the results from a Draft Electric Power

Research Institute (EPRI) Report TR-103680, Rev. 1, November 1995,

``Empirical Study of Iodine Spiking In PWR Power Plants'' were

applied. The objective of the EPRI study was to quantify the iodine

spiking in a postulated Main Steam Line Break/Steam Generator Tube

Rupture (MSLB/SGTR) accident sequences. In the EPRI report, an

iodine spike factor between 40 and 150 was determined to match data

from existing plant trips. The maximum iodine spike factor value of

150 was applied to a steady-state equilibrium RCS DE I-131 activity

of 0.33 microCuries/gram. The resulting two-hour average iodine

concentration for a postulated MSLB/SGTR accident sequence was

determined to be 3.1 microCuries/gram. Since the EPRI report is

based on industry data and the EPRI method predicted a post-accident

iodine activity, which is a small fraction of the activity predicted

by the NRC SRP methodology, it can be expected that, for the

proposed 0.05 microCuries/gram limit under an MSLB/SGTR accident

sequence, the post-accident iodine activity would typically be a

small fraction of the RCS DE I-131 activity predicted by the NRC SRP

methodology. For Braidwood, using the SRP methodology with an RCS DE

I-131 activity of 1.0 microCuries/gram and a spike factor of 500,

the Post-Trip RCS activity two hours after the event would be near

38 microCuries/gram. At an RCS DE I-131 activity of 0.05

microCuries/gram, it would require a spike factor of nearly 10,000

to obtain a Post-Trip RCS DE I-131 activity near 38 microCuries/

gram. With a Post-Trip RCS DE I-131 activity of 38 microCuries/gram,

an increase in the allowable leak rate could impact the 10 CFR 100

limits. To accommodate for an increase in the allowable leak rate by

a factor of twenty, the resultant activity would need to be below

1.9 microCuries/gram. Two (2) of the seventeen (17) post-trip data

points from Braidwood exceeded 1.9 microCuries/gram. Both occurred

during cycles with fuel defects. Braidwood Unit 1 is currently

operating with no fuel defects. The conservatisms mentioned below

will reduce the possibility of exceeding a small fraction of the 10

CFR 100 limits should the post-trip iodine exceed 1.9 microCuries/

gram.

Based on evaluations by the four methods above, Braidwood can

conclude that the current methodology (Method 1) used to predict

iodine spiking is conservative. Although dose projections indicate

with confidence that the iodine spiking factor limit will be met,

the conservatisms in the offsite dose calculation and current

Braidwood Unit 1 operating conditions listed below, provide added

assurance that the 10 CFR 100 limits, General Design Criteria (GDC)

19 criteria, and the requirements of NRC Generic Letter 95-05 will

be satisfied if the iodine spike factor exceeds 500 or the post-trip

fuel release rate exceeds 686.7 Ci/hr.

As further assurance that the 10 CFR 100 and GDC 19 limits are

not exceeded, several conservatisms are inherent to the offsite dose

calculation. These conservatisms include, but are not limited to:

1. The meteorological data used is at the fifth percentile. It

is expected that the actual dispersion of the iodine would result in

less exposure at the site boundary than the 30 Rem limit of 10 CFR

100.

2. Iodine partitioning is not accounted for in the faulted SG.

With the high pH of the secondary water, some partitioning is

expected to occur. An iodine partition factor of 0.1 is more

realistic (per Table 15.1-3 of Reference 8 [the Braidwood Updated

Final Safety Analysis Report]) than the 1.0 valued (no partitioning)

used in the offsite dose calculation. This reduces calculated dose

by 90%.

3. The activity in the RCS is not expected to increase

instantaneously with the spike in iodine released from the defective

fuel.

4. The results from the Braidwood tube pull data indicate that

the projected Interim Plugging Criteria leak rate is conservative.

In addition, the current Braidwood Unit 1 operating conditions

provide defense in depth and provide further assurance that the 10

CFR 100 and GDC 19 limits will not be exceeded:

1. Braidwood Unit 1 is currently operating with a debris

resistant fuel design which is less likely to develop fuel defects.

2. As evidenced by industry data, if debris related fuel

failures are going to occur they are most likely to be occur early

in the cycle. Braidwood Unit 1 has operated approximately 6 months

into its current cycle and has seen no signs of fuel defects.

Therefore, fuel failure prior to completion of the current cycle is

not likely.

3. The RCS DE I-131 activity is likely to be less than the TS

limit. With the current Braidwood Unit 1 RCS DE I-131 activity near

3E-4 microCuries/gram with no fuel defects, the spike factor is

expected to be considerably smaller than the 500 value.

4. It is unlikely, for the short time period this amendment is

being requested (remainder of Cycle 7), that an accident-initiated

iodine spike for Braidwood Unit 1 would be greater than the NRC SRP

assumed value.

5. Primary-to-secondary leakage is likely to be less than the TS

limit (150 gpd) in each of the four SGs prior to the event.

Currently, minimal primary-to-secondary leakage (less than 5 gpd)

exists at Braidwood Unit 1.

These proposed changes do not result in a significant increase

in the consequences of an accident previously analyzed.

The RCS DE I-131 activity limit is not considered as a precursor

to any accident. Therefore, this proposed change does not result in

a significant increase in the probability of an accident previously

analyzed.

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

or different kind of accident from any accident previously

evaluated.

The changes proposed in this amendment request conservatively

reduce the Unit 1 RCS DE I-131 activity limit at which action needs

to be taken. The changes do not directly affect plant operation.

These changes will not result in the installation of any new

equipment or systems or the modification of any existing equipment

or systems. No new operating procedures, conditions or

configurations will be created by this proposed amendment.

Accordingly, this proposed change does not create the

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

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

NRC Generic Letter 95-05 allows lowering of the RCS dose

equivalent iodine as a means for accepting higher projected leakage

rates provided justification for the RCS DE I-131 activity below

0.35 microCuries/gram is provided. Four methods for determining the

fuel rod iodine release rates and spike factors during an accident

were reviewed. Each of these methods utilized actual industry data,

including Braidwood Units 1 and 2, for pre-and post-reactor trip RCS

DE I-131 activities. Each of the methods demonstrated that the

actual fuel rod iodine release rates are a small fraction of the

release rate as calculated using the NRC SRP methodology. Although

these values are a small fraction of that determined by the NRC SRP

Method, Braidwood is also requesting an increase in the allowable

primary-to-secondary leak rate during MSLB. By decreasing the TS RCS

DE I-131 activity limit by a factor of twenty and increasing the

allowable leak rate by a factor of twenty, the activity released to

the public would be equal to or less than the activity calculated by

the SRP method for each of the seventeen reactor trip events

reviewed at Braidwood. The predicted end-of-cycle 7 leak rate is

122.3 gpm (Room T/P [temperature and pressure]). The calculated site

boundary dose due to this leakage is 27.63 Rem. This dose meets the

requirements of 10 CFR 100 and GDC 19. All design basis and off-site

dose calculation assumptions remain satisfied. This proposed change

would not result in a 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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Wilmington 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 60603.

[[Page 11917]]

NRC Project Director: Robert A. Capra.

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: September 24, 1997.

Description of amendment request: The proposed amendment would

revise Technical Specification Surveillance Requirement 4.3.4.2 to

change the frequency of turbine throttle and governor valve testing

from monthly to quarterly and incorporate corresponding administrative

changes. Bases 3/4.3.4 will be changed to update a referenced vendor

document and incorporate corresponding administrative changes.

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 Bases change is a reference update, which is administrative

in nature. Additional administrative changes necessitated by a

change in the presentation of the surveillance requirements are

proposed. The changes are consistent with Generic Letter 93-05 and

NUREG-1366. This change reduces the frequency of testing that is

likely to cause transients or excessive wear of equipment. An

evaluation of these changes indicates that there will be a benefit

to plant safety. The evaluation, documented in NUREG-1366,

considered (1) unavailability of safety equipment due to testing,

(2) initiation of significant transients due to testing, (3)

actuation of engineered safety features that unnecessarily cycle

safety equipment, (4) importance to safety of that system or

component, (5) failure rate of that system or component, and (6)

effectiveness of the test in discovering the failure.

As a result of the decrease in the testing frequencies, the risk

of testing causing a transient and equipment degradation will be

decreased, and the reliability of the equipment will not be

significantly decreased.

The initial conditions and methodologies used in the accident

analyses remain unchanged. The proposed changes do not change or

alter the design assumptions for the systems or components used to

mitigate the consequences of an accident. Therefore, accident

analyses results are not impacted. Appropriate testing will continue

to assure that equipment and systems will be capable of performing

the intended function. The frequency of testing is not a precursor

for any analyzed accidents.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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 modify allowable intervals between turbine

throttle and governor valve surveillance tests. The proposed changes

do not affect the design or operation of any system, structure, or

component in the plant. The safety functions of the related

structures, systems, or components are not changed in any manner,

nor is the reliability of any structure, system, or component

reduced by the revised surveillance or testing requirements.

Appropriate testing will continue to assure that the system is

capable of performing its intended function.

The changes do not affect the manner by which the facility is

operated and do not change any facility design feature, structure,

system, or component. No new or different type of equipment will be

installed.

The turbine valve testing surveillances will be changed to

account for a frequency change from monthly to quarterly for the

throttle valves and for the governor valves.

Since there is no change to the facility or operating

procedures, and the safety functions and reliability of structures,

systems, or components are not affected, the proposed changes do not

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

any accident previously evaluated.

C. The proposed changes do not involve a significant reduction

in a margin of safety.

All of the proposed Technical Specification changes are

compatible with plant operating experience and are consistent with

the guidance provided in Generic Letter 93-05 and NUREG-1366. The

changes reduce the frequency of testing that increases the risk of

transients and equipment degradation. There is no impact on safety

limits or limiting safety system settings. The Bases change is a

vendor reference update, which is administrative in nature.

Certain reload designs can be such that power differences

between the top and bottom of the core are more sensitive to control

and can develop divergent xenon oscillations when the power

reduction occurs during the middle of core life. Near the end of

core life, stabilizing even larger differences in axial power

distribution becomes more of a problem because of the larger

temperature coefficient, lower boron concentration and larger

differential xenon transient. In the Safety Evaluation Report

related to the Prairie Island Amendment Numbers 86 and 79 in regard

to the discussion above, the NRC wrote, ``Based on the above, the

staff has concluded that the margin of safety is reduced when the

plant is undergoing turbine valve testing.''

Since this amendment reduces the number of turbine tests while

still maintaining acceptable equipment reliability, the proposed

changes result in an increase 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 10 CFR 50.92(c) are

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

requested amendments involve no significant hazards consideration.

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

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

for Braidwood, the Wilmington 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 60603.

NRC Project Director: Robert A. Capra.

Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power

Plant, Unit 1, Monroe County, Michigan

Date of amendment request: December 15, 1997 (Reference NRC-97-

0115).

Description of amendment request: The proposed amendment will

revise License Condition A to delete references to letters dated May

17, 1985, July 23, 1986, September 15, 1986, September 25, 1987,

September 15, 1988, and December 22, 1988, and replace them with the

Enrico Fermi Atomic Power Plant, Unit 1, Safety Analysis Report (F1SAR)

as the licensing basis.

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 using the standards in 10 CFR 50.92(c). The licensee's

analysis is presented below:

(1) Does the proposed change significantly increase the

probability or consequences of an accident previously evaluated?

No, the proposed submittal of the F1SAR as the facility's

licensing basis document does not significantly increase the

probability of an accident. The F1SAR is a compilation of previously

submitted information and other information gathered on the

condition of the facility. Compilation of current information and

imposition of the new Fire Protection and Quality Assurance Program

requirements will not increase the probability of an accident. These

additional controls would reduce the probability of an event. The

proposed addition of a hypothetical secondary sodium accident

scenario identifies one possible previously unidentified potential

cause of a primary sodium release and/or liquid waste tank release.

The previous submittal assumed the cause of the primary sodium

release to be a fire or other catastrophic event. The cause of the

liquid waste tank rupture was assumed to be an earthquake.

Recognition of a cause being the reaction of secondary sodium does

[[Page 11918]]

not significantly increase the probability of a primary sodium

release or liquid waste release. A catastrophic event would still

need to occur to cause the postulated scenario, so there is no

discernible increase in the probability of the primary sodium or

liquid waste accident compared to the existing licensing basis. For

the reasons discussed above, substituting the F1SAR as the licensing

basis for Fermi 1 will not significantly increase the probability of

an accident.

The proposed submittal of the F1SAR as the Fermi 1 licensing

basis document will have no impact on the consequences of an

accident. Consolidating current information on the plant and

previous submittals does not change the amount of radioactivity at

the facility or the potential magnitude of any release during an

accident. Since the potential accident source terms were not updated

as part of the submittal, the consequences of the accidents

contained in the F1SAR match the consequences in the previous

submittal. Though a new postulated hypothetical accident scenario

was added, the secondary sodium involved in that accident is not

radioactive, per previous submittals, and so the only potential

radiological consequences of that scenario occur if the primary

sodium or liquid waste is released and those consequences have

already been reviewed in the NRC safety analysis for Amendment No. 9

to the Fermi 1 license. Therefore, the adoption of the F1SAR as the

facility's licensing basis will not significantly increase the

consequences of an accident at Fermi 1.

(2) Will the proposed amendment create the possibility of a new

or different kind of accident from any accident previously analyzed?

No, establishment of the F1SAR as the Fermi 1 licensing basis

document will not create a new type of accident. The F1SAR is mainly

a compilation of the previous licensing basis documents, information

on the facility condition and additional controls. It does not

involve operating in any new type of mode and so cannot create a new

or different type of accident. The new hypothetical secondary sodium

accident contained in the F1SAR is a sodium accident. One of the

existing licensing basis accidents is the primary sodium accident

resulting in release of the primary sodium and its activity. The

hypothetical secondary sodium accident as analyzed may lead to the

release of the primary sodium or liquid waste and so it is a

potential precursor of an already identified accident.

(3) Will the proposed change significantly reduce the margin of

safety at the facility?

No, adopting the new F1SAR as the licensing basis document for

Fermi 1 will not decrease the margin of safety. It will establish an

up-to-date licensing basis, so future changes can be appropriately

evaluated against an updated safety analysis report. The F1SAR

better describes the current condition of the plant. No physical

changes will be implemented based on the submittal of the F1SAR.

Some additional administrative requirements will be established in

the new Quality Assurance program and in the need to keep the F1SAR

updated biannually. No new types of accidents are discussed in the

F1SAR--the discussion of the hypothetical secondary sodium event is

a more detailed discussion of what potentially could happen during a

catastrophic event leading to a sodium reaction. A total primary

sodium release was already established as a licensing basis event.

Because the F1SAR will not, in itself, lead to physical changes, but

will be the new standard to which future changes are compared,

establishment of this updated document as the Fermi 1 licensing

basis will not significantly reduce the margin of safety of the

facility.

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, NRC staff proposes to determine that the amendment request

involves no significant hazards consideration.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

Attorney for licensee: John Flynn, Esquire, Detroit Edison Company,

2000 Second Avenue, Detroit, Michigan 48226.

NRC Branch Chief: John W. N. Hickey.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver

Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

Date of amendment request: December 19, 1997.

Description of amendment request: The proposed amendment would

revise the requirements for the source range neutron flux channels in

Mode 2 (Below P-6), 3, 4, and 5 to incorporate the guidance provided in

NUREG-1431, the NRC's Improved Standard Technical Specifications (ISTS)

with some modifications to address plant-specific design features. This

change would allow (1) the use of alternate detectors provided the

required functions are provided, and (2) plant cooldown with inoperable

detectors provided the shutdown margin accounts for the temperature

change. This change would also modify the Unit 2 Technical

Specifications (TS) Table 3.3-1 Channels To Trip and Minimum Channels

Operable requirements to 0 and 1, respectively. This portion of the

amendment would make these Unit 2 requirements consistent with the

current Unit 1 requirements. For both Units 1 and 2, TS Table 4.3-1

would be modified to include a notation exempting the alternate source

range detectors from surveillance testing until they are repaired for

operability.

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. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed amendment would modify the reactor trip system

instrumentation requirements to permit the use of alternate

detectors in place of inoperable source range detectors. The

alternate detectors will be connected to the source range circuits

to provide the required indications and functions. The alternate

detectors are not required to be tested to satisfy the surveillance

requirements until they are connected to the source range circuits

and required to be operable. The alternate detectors must have the

accuracy and sensitivity required to adequately monitor changes in

the core reactivity levels. The alternate detectors will provide

neutron flux monitoring in place of the source range detectors thus

assuring core monitoring at a level consistent with the current

technical specification requirements. Therefore, there is no loss of

function or need for additional compensatory actions and the

operators can perform required plant evolutions while relying on the

alternate detectors.

Two operable detectors are required when the control rods are

capable of withdrawal. Rod withdrawal and boron dilution add

positive reactivity which can significantly affect the reactivity

condition of the core, therefore, two monitors are required operable

during startup evolutions. Redundant detectors are required to

ensure that two source range neutron flux detectors are available to

detect changes in core reactivity. These changes provide those

indications and functions consistent with the current technical

specification requirements where at least two source range detectors

are operating and capable of providing the required functions. The

function of the source range detectors is to provide direct neutron

flux monitoring of the core to detect changes in reactivity which

would result in a loss of the required shutdown margin.

One source range or alternate detector is required when the

control rods are fully inserted and are not capable of withdrawal.

Plant cooldown is recognized as a positive reactivity addition,

however, this is accounted for in the shutdown margin calculations.

The shutdown margin remains essentially unchanged and will be

available to preclude a criticality event during this evolution.

Inadvertent control rod withdrawal is not a concern, therefore, one

source range or alternate detector can adequately monitor the core

neutron flux. The action statements have been modified to address

the NUREG-1431 Improved Standard Technical Specification (ISTS)

requirements along with incorporating the ability to use alternate

detectors in place of the source range detectors.

Bases 3/4.3.1 and 3/4.3.2, Protective and Engineered Safety

Features (ESF) Instrumentation, has been revised to include the

modifications to the source range detector requirements including

the use of alternate

[[Page 11919]]

source range detectors. The alternate detectors must provide

sufficient accuracy and sensitivity to adequately monitor changes in

core reactivity during Modes 2 (Below P-6), 3, 4, and 5.

The operability requirements of the source range neutron flux

instrumentation will continue to be met when using an alternate

detector in place of a source range neutron flux detector. No

changes are being incorporated that would act to increase the

probability of a positive reactivity addition event, therefore, the

proposed change will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The function of the source range detectors is to provide direct

neutron flux monitoring of the core to detect positive reactivity

additions which would result in a loss of the required shutdown

margin. The alternate detectors must provide the accuracy and

sensitivity required to adequately monitor changes in the core

reactivity levels during shutdown and startup activities. The

alternate monitors will be connected to the source range circuits to

provide the required indications and functions. Therefore, there is

no loss of function or need for additional compensatory actions and

plant shutdown and startup activities can be continued while relying

on the alternate detectors.

Control rod withdrawal is a method capable of providing rapid

positive reactivity addition with boron dilution being a much slower

positive reactivity addition method. With the control rods capable

of withdrawal, a rod withdrawal event could rapidly initiate core

criticality so redundant source range detectors are required

operable. This ensures adequate monitoring capability is available

to alert the operators of a rapid increase in the core reactivity

condition. The maximum reactivity addition due to the boron dilution

is slow enough to allow the operator to determine the cause and take

corrective action before the shutdown margin is lost. These changes

will not affect the operability or reliability of the source range

instrumentation to provide the required indications and functions.

Therefore, the proposed change will not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. Does the change involve a significant reduction in a margin

of safety?

The proposed change will continue to ensure the required source

range instrumentation functions are available during shutdown and

startup conditions. This change will not reduce the reliability of

the source range detectors to monitor the core reactivity condition

and provide the appropriate indications or affect the required

shutdown margin. Plant operation will continue to be maintained

within the shutdown margin requirements of [Technical] Specification

3.1.1.1 and 3.1.1.2. The required indications and functions are

still maintained in accordance with current technical specification

requirements and the shutdown margin is unaffected, therefore, the

proposed change will 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.

Local Public Document Room location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA 15001.

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

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

NRC Project Director: John F. Stolz.

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power

Station, Unit 2, Shippingport, Pennsylvania

Date of amendment request: January 29, 1998.

Description of amendment request: The proposed amendment would

revise the Beaver Valley Power Station, Unit No. 2, Updated Final

Safety Analysis Report (UFSAR) calculated doses to address a non-

conversative assumption regarding control room emergency pressurization

fan flow during the Locked Rotor accident and include new X/Q values in

calculating the Exclusion Area Boundary (EAB) and Low Population Zone

(LPZ) doses.

This change is not the result of hardware changes to the plant or a

change in operating practices. It reflects corrected analysis results

only and allows correction of the licensing basis to reflect

conservative assumptions used in the revised dose analysis for a Locked

Rotor event.

The proposed amendment would also revise USFAR Tables 15.0-13,

15.6-15 and 15.6-16 to modify calculation parameters and UFSAR Section

15.6.5.5 to include editorial changes to ensure that descriptions of

the Small Break Loss of Coolant Accident (SBLOCA) radiological

consequences are clear. The following items in the UFSAR description of

the SBLOCA radiological consequences analysis were changed: (1) a new

lower minimum control room emergency pressurization fan flow rate and

(2) a new lower minimum air bottle discharge rate.

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. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

[Locked Rotor Accident]

The proposed amendment would revise the calculated control room

doses for a Locked Rotor accident to address a non-conservative

assumption for the fan pressurization system flow rate. The proposed

amendment does not affect the capability of the control room

habitability system to maintain control room dose within the limits

of General Design Criterion (GDC) 19 in Appendix A of the Code of

Federal Regulations Title 10 Part 50. The control room habitability

system is an accident mitigation system and will continue to operate

as designed. The system has no accident prevention function nor does

it interact with systems that have such a function. The proposed

change does not alter plant systems, structures or components.

The proposed amendment would also revise calculated offsite

doses resulting from a locked rotor accident. This change in doses

is not due to physical plant changes, but results mainly from use of

more conservative assumptions used in calculating doses.

The proposed change does not affect the manner in which the

plant is operated. The physical plant equipment and operating

practices are not changed; therefore, the probability of an accident

previously evaluated remains unchanged.

The performance requirements of the plant systems which are

required to minimize the radiological consequences of a Locked Rotor

accident remain unchanged. The proposed change slightly increases

calculated control room doses due to an analysis input change for

filtration fan flow rate. This slight increase remains below the

limits required by GDC 19. The proposed change does not involve a

significant increase in the consequences of an accident previously

evaluated since adequate control room radiation protection continues

to be provided to ensure actions can be taken to operate the plant

safely under accident conditions. The radiological consequences to

the environment from a Locked Rotor accident remain unchanged since

the performance of plant systems remains unchanged. Although

slightly increased, revised calculated offsite doses remain less

than 10 CFR 100 limits.

[SBLOCA]

The proposed amendment would revise the control room dose

analysis parameters for a Small Break Loss of Coolant Accident

(SBLOCA) to include more conservative assumptions for the

pressurization system flow rate. The proposed amendment does not

affect the capability of the control room habitability system to

maintain control room dose within the limits of General Design

Criterion (GDC) 19 in Appendix A of the Code of Federal Regulations

Title 10 Part 50. The control room habitability system is an

accident mitigation system and will continue to operate as designed.

The system has no accident prevention function nor does it interact

with systems that have such a function. The proposed change does not

alter plant systems, structures or components.

The proposed change does not affect the manner in which the

plant is operated. The physical plant equipment and operating

[[Page 11920]]

practices are not changed; therefore, the probability of an accident

previously evaluated remains unchanged.

The performance requirements of the plant systems which are

required to minimize the radiological consequences of a SBLOCA

remain unchanged. The proposed change slightly decreases calculated

control room doses due to analysis input changes. Calculated doses

remain below the limits required by GDC 19.

Based on the above discussion, it is concluded that th[e]

proposed change[s] [do] not involve a significant increase in the

probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

[Locked Rotor Accident]

The proposed change does not alter the method of operating the

plant nor does it pose additional challenges to the design or

function of the control room habitability system. The control room

habitability system will continue to operate as designed. The

control room habitability system will continue to maintain the

control room dose consequences within the limits specified in GDC

19. Adequate control room radiation protection will continue to be

provided to ensure actions can be taken to operate the plants safely

under accident conditions. The proposed change to the control room

dose is only the result of a change in analysis input parameters.

Plant performance has not been modified in any way which affects

doses to the public.