ML20211A681
| ML20211A681 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 09/17/1997 |
| From: | Woodard J SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20211A685 | List: |
| References | |
| GL-95-05, GL-95-5, NUDOCS 9709240330 | |
| Download: ML20211A681 (14) | |
Text
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- s. ,1 D.Wooderd Southern Nucle:r Operating Compery,1:c.
tucutive We Presdent 4 invemes: Center Parkway PO Ban 1795 Dirmingham. Alabama 35201 Tel 205 99? $3B6 September 17. 1997 SOUTHERN COMPANY Energy to Serve YourWorld" l
Docket No.: 50 348 10 CFR 50.90 50 364 U.S. Nuclear Regulatory Commission A1TN: Document Control Desk Washington, DC 20555 Joseph M, Farley Nuclear Plant Primary CoohtrlLSpecific Activity Technicadpecification Amendment Ladies and Gentlemen:
13y letter dated August 29,1997, Southern Nuclear fonvarded the Joseph M. Faricy Nuclear Plant, Unit 1, 90 day report concerning operation with the alternate repair criteria for ODSCC at tube support plate intersections. While that report contained infornation on the tube pulled in the previous Unit I steam generator inspection, it did not provide the results of updated calculations since they had not been completed, llowever, the cover letter for the report indicated that including the new data point in the database would have a significant impact on calculations associated with the alternate repair criteria.
In a conference call with the NRC Stafron September 11,1997, Southern Nuclear infomied the Staff that the inclusion of the data point in the data base would result in a probability of burst in the event of a steam 4
line break of 1.2 x 10 and projected primary-to-secondary leakage in the event of a steam line break of 20.4 gpm. The density compensated steam line break leakage limit for Farley Unit I is 13.7 gpm.
Enclosure i contains the Generic Letter 95-05 Safety Assessment provided to the NRC StafTon September 12,1997 documenting the acceptability of the probability of burst during a steam line break and acceptability of the steam line break primary-to-secondary steam generator leakage based on the implementation of compensatory measures.
To insure the doses at the site boundary do not exceed the regulatory limits if a steam line break occurs, the dose equivalent iodine on Unit I has been limited to 0.15 Curies / gram for steady state conditions. The transient limit has also been appropriately reduced, for example, a limit of 9 pCuries/ gram for power levels above 80%. Although Unit 2 calculations indicate that currently projected leakage in the event of a steam line break is acceptabic, it is projected to exceed the limit on approximately November 6,1997.
Consequently, the Unit 2 dose equivalent iodine limit will be administratively reduced to match the Unit I limits by November 1,1997.
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o U. S. Nuclear Regulatory Commission Page 2 A technical specification amendment for a reduction in the specific actisity limits of the Farley Units I and 2 technical specifications is submitted. Enclosure 2 provides the basis of the described amendment.
Enclosure 3 provides a significant hazards evaluation for the amendment. Enclosure 4 provides the revisal technical specification pages.
Southern Nuclear has performed an assessment of the impact of the proposed revision to the technical specifications on the environment and has determined that there is no impact on the enviromnent. %c proposed revision does not affect the types or amounts of any radiological or non radiological effluents that may be released offsite. No increase in individual or cumulative occupational radiation exposures will iesult from this revision. Additionally, the revision does not involve the use of any resources not previously considered in the Final Environmental Statement related to the operation of Failey Nuclear Plant.
l A copy of these proposed changes is being sent to Dr. D. E. Williamson, the Alabama State Designec, in accordance with 10 CFR 50.91(b)(1).
Southern Nuclear will maintain the above administrative limits until the NRC StafTapproves the proposed specific activity limits amendments; until alternative calculation methods are approved; or until the steam generators are replaced. Approval of this technical specification is requested on an expedited basis.
If you have any questions, please advise.
Respectfully submitted, QUTilERN NUCLEAR OPERATING COhtPANY W O's J. . Woodard M or omd subscribed before me thi./.20tay o(yI991 h! . kd ublic hty C mmission Expires: f-M ~78 i
REh1/maf:15 del. DOC
Enclosures:
- 1. Generic Letter 95 05 Safety Assessment
- 2. Basis for Amendment
- 3. Significant flazards Evaluation
- 4. Technical Specification Pages cc: hir. L. A. Reyes, Region 11 Administrator hir. J.1. Zimmerman, NRR Project hianager hir. T. hl. Ross, Plant Sr. Resident inspector Dr. D. E. Williamson, State Department of Public Ilealth
4 Enclosure i Generic Letter 95-05 Safety Assessment
Revision 4 GENERIC LETTER 95 05 SAFETY ASSESSMENT
))KKGRol/ND Farley Nuclear Plant Unit 1 is currently licensed to use the Alternate Repair Criteria (ARC) for ODSCC at tube support plates. The permanent ARC for Unit I was approved by license Amendment 124. The ARC is based on guidance contained in Generic Letter 95-05. This generic letter states in part that 1) the calculatbns of steam line break (SLB) steam generator leakage and conditional probability of steam generator tube burst should be conducted in accordance with the guidance of the generic letter and 2) that tubes must be periodically removed from the SG's, examined, and destructively tested to verify the morphology of the degradation and provide additional data for the structural and leakage evaluations.
Recently an industry issue arose that impacts the SLB leakage limit and calculated EOC SLB l steam generator leakage. (Re: NEl letter dated July 30,1997) The issue relates to comparing i site allowable SLB steam generator leak f ates to predicted EOC SLB leak rates that have not been density compensated. FNP has evt,uated this issue for impact on the Unit 1 ARC leak rate.
In this evaluation it was found the pt.oously calculated SLB steam generator leakage limit was reduced from 19 gpm to a density compensated limit of 13.7 gpm. The maximum projected EOC leak rate for the current operating cycle, based on the last SG inspection was 15.7 gpm. Thus, the projected EOC leakage exceeds the limit. This was reported to the NRC in the Unit i 90 day ARC report submitted on August 29,1997. In addition, an evaluation was conducted to l determine when the calculated SLB steam generator leak raic would exceed the 13.7 gpm limit.
This was determined to be approximately 310 EFPD into the cycle. (Note: As of September 12, 1997 Unit I was 100 days into the cycle.)
During the spring 1997 Unit 1 RF 14 outage a 13.7 volt indication was found in the IC SG at the first tube support plate (TSP). This was considered to be a significant fmding with regard to the existing data base for TSP ARC. This tube was in situ tested and found to have zero leakage at steam line break (SLB) differential pressure. The tube was subsequently pulled, tested, and destructively examined. The SLB leakage for this intersection was found to be 1641/hr (0.72 gpm) and the burst pressure was 3599 psi normalized for nominal material properties. When this data is added to the currently approved voltage versus burst correlation and the voltage versus SLB leakage correlation, there is a significant impact on the projected calculated EOC leakage. ,
The new EOC calculated leak rate was determined to be 20.4 gpm. The new tube data also l
affected the calculated probability of burst (POB) value. This value increased from 9.9x10 to 1.2x104.
Farley Nuclear Plant will submit a Technical Specification amendment to ensure doses at the site boundary in the event of a SLB will not exceed the regulatory limits. The Technical Specification amendment will be submitted by September 17,1997.
Generic Letter 95-05 Safety Assessment Page 2
/*0TliNTIAI, NON-CONFORMING CONDITION Using the current licensing basis data base (excluding the latest Farley Units 1 and 2 pulled tubes) for burst and leak rate correlations, Unit 1 is within the density compensated 13.7 gpm limit and 2
the lx10 P0H limit. Unit I will remain within the 13.7 gpm limit until approximately 310 EFPD into the cycle. Ilowever, using the new burst and leak rate correlations updated with the FNP 13.7 volt data, FNP Unit I calculated SLB leakage exceeds the density compensated 13.7 gpm leakage limit and the probability of burst exceeds the lx10 limit.
Ol'liRAllis ITY ASSliSSMiiNT l Generic Letter 95 05 requires that "A licensee that chooses to implement these criteria should l include all of the following in the proposed program:... Reporting of all results according to the guidance discussed in Section 6 of Attachment 1."
In the application for amendment for ARC dated 12 6 96, FNP made a commitment to follow GL 9* 05, in this letter SNC committed to follow Section 6 of the GL regarding staff notification to I continue using the ARC. Specifically, Southern Nuclear committed that "Results will be reported per the guidance of Section 6 of Attachment 1 of Generic Letter 95 05." Southern Nuclear committed to follow the reporting requirements of Generic Letter 95 05 as discussed below.
FNP is currently licensed for use of the ARC for indications at tube suppon plate intersections per Technical Specification (T.S.) 3.4.6. In the approved 13ases for T.S. 3.4.6 it states that the voltage-based repair limits of T.S. 4.4.6 implement the guidance in GL 95-05. In the NRC issued SER for Amendment 124, which approved the ARC,it states, GL 95 05 specifies that the structural and leakage integrity assessments should use the latest data from destructive examinaiions of tubes removed from licensees' steam generators. The licensee stated that the latest NRC-approved database, using NRC approved data exclusion criteria, will be applied to the tube integrity evaluations. For the upcoming Parley Unit 1 inspection and GL 95 05 specified calculations, the licensee will use the database forwarded to the NRC by Duquesne Light Company for Beaver Valley Unit 1, dated March 27,1996. The database contains the most currently available tube pull data from industry and also satisfies the exclusion criteria specified in GL 95 05. Therefore, the stafTfmds that the database submitted by Duquesne Light Company is acceptable for the GL 95-05 calculations for the upcoming Farley Unit 1 inspection.
1 The SER goes on to amplify how the database should be updated by stating:
For the long-term, Nuclear Energy Institute has developed a protocol for updating the steam generator degradation database. The staff will review the adequacy of the protocol Pending the implementation of an NRC approved process for updating a generic industry database for steam generator tube degradation, the licensee will provide the NRC with the database it intends to use prior to each refueling outage. The database will include the data from tubes that have been
Generic Letter 95-05 Safety Assessment Page 3 pulled and tested up to 2 months before the plant outage. The staff finds the licensee's proposal acceptable.
The above clearly shows how FNP is committed to follow the guidance of GL 95 05. With regard to how to deal with new data from pulled tubes and their effect on the burst and leak rate calculations Section 6 of the GL states:
- 6. Reporting Requirements 6.a Threshold Criteria for Requiring Prior StafTNotification To Continue With Voltage Based Criteria This guidance allows licensees to implement the voltage based repair criteria on a continuing basis after the NRC staff has approved the initial TS amendment, llowever, in several situations, the NRC staffmust receive notification to enable l the staff to assess whether a licensee can continue with the implementation of the l voltage-based repair critetia:
1 6.a.1 If the projected EOC voltage distribution results in an estimated leakage greater than the leakage limit (determined from the licensing basis calculation),
j then the licensee should notify the NRC of this occurrence and provide an assessment ofits significance prior to returning the SGs to service. Ifit is not practical to complete this calculation prior to returning the SGs to service, the measured EOC voltage distribution can be used (from the previous cycle of operation) as an alternative (refer to Section 2.c). Ifit is determined that the projected calculated leakage will exceed the leakage limit (during the operating cycle) after the SGs are returned to service, then lleensees should g
provide an assessment of the safety significance of the occurrence, describe the compensatory measures being taken to resolve the issue, and follow any other applicable reportability regulations.
When FNP determined that the projected Unit I leak rate could exceed the limit when the new 13.7 volt data was added to the database, a safety assessment was conducted and notification of the NRC was set up. The NRC was notified on a phone call on September 11,1997 of the impact of the addition of the tube to the data base and the density compensation issue and their effect on meeting the SLB leakage limit and POB limit. The safety assessment and compensatory measuret listed below are provided to the staff.
Thus, FNP has complied with the requirements of Technical Specification 3.4.6 by having ,
complied with Section 6 of GL 95-05 as described above and the SG's are considered operable.
4 Gen:ric Letter 95 05 Safety Assessment Page 4 SAFETYASSESSMENT During normal power operations, dose equivalent iodine at Farley Unit I remains less than 0.01 pCuric/ gram. Dose equivalent lodine has not exceeded 0.15 pCurie/ gram since the startup from the last refbcling outage. At these dose equivalent lodine levels, doses at the site boundary in the event of a steam line break would have been below the regulatory limits.
A Technier! Specl0 cation amendment to lower the dose equivalent lodine limits will be submitted by September 17,1997.
With the 13.7 volt tube included in the burst data, the probability of burst would be 1.2 x 10,
which is slightly above the i x 10 limit required by Generic Letter 95-05. The value of 1.2 x 10 2 probability of burst is calculated assuming RCS pressure rises to 2560 psi after every steam line break. Given a steam line break accident, emergency operating procedures instruct operators to
- terminate the safety injection and limit the RCS pressure rise. The Parley PRA utilizes a l probability ofless than 10 that operators would fail to terminate the safety injection. This is substantiated by PRA analysis and observations of numerous crews on the Farley simulator.
Thus, the actual likelihood of reaching 2560 psi on every steam line break is lower than uti;ized in the Generic Letter 95 05 calculation for probability of burst. Using engineeringjudgment, the conditional probability of burst when accounting for operator actions to limit the pressure following the steam line break would reduce the 1.2 x 102 probability to below the limit of 1.0 x 10' .
COMPENSATORYMEASURES Farley Nuclear plant will administratively limit the dose equivalent iodine to 0.15 pCurie/ gram steady state and appropriately reduce the limit for transient conditions (for example, a limit of 9 pCurie/ gram at power levels above 80%). As a result of this reduction in dose equivalent iodine limits, the revised primary to secondary leakage limit in the event of a steam line break will be 27.6 gpm (density compensated). The projected primary-to secondary leakage for a steam line break occurring at the end of the current operating cy:lc on Farley Unit 1 is 20.4 gpm, well below the revised limit.
Farley Nuclear Plant will submit a Technical Specification amendment to ensure doses at the site boundary in the event of a SLB will not exceed the regulatory limits. The Technical Specification amendment will be submitted by September 17,1997.
A DIblTIONAI. SUPlwRTING INFORMA TION _ ;
Generic Letter 95 05 provides specific guidance on performing calculations to project the steam '
line break induced leakage rate. As a result of the specific guidance, a number of conservatisms are included in the calculation.
If the linear regression fit is not valid at the 5% level with a "p value" test, the linea; cegression fit that is used for the simulation is assumed to have a zero slope. In other words, all flaws are
< assumed to leak at the same rate, regardless of bobt in voltage. Folloiving inclusion of the most recent Farley Unit I and 2 pulled tube results, the correlation ofleak rate to bobbin voltage um
O Generic Letter 95-05 Safety Assessment Page5 exhibits a p-value of 7.6%. If the French data is excluded fr :n the database, e.g., only the Farley
_ pulled tubes and model boiler specirrens are used, the corr .ation exhibits a p value of 2.0%. To demonstrate the impact of this requirement, the followb sble is provided.
Methodology Leakage Projection Currently approved methodology (WCAP-14277),
60% probability of detection, 20.4 gpm zero slope leak rate regression correlation, all currently available data in database (French data and latet Farley pulled tubes)
Currently approved methodology (WCAP-14277),
60% probability of detectiw., 3.8 gpm l leak rate correlation based on use of Farley pulled tube data and model boiler specimens (French data deleted)
Currently approved methodology (WCAP-14277),
60% probability of detection, 3.1 gpm leak rate correlation based on use of Parley pulled tube data, French data, and model boiler specimens less MB-542 4 It should be explained that the two data points which preclude meeting the 5% p value test are a 4 French data point with a bobbin voltage of 30.9 volts which leaked at a rate of 0.13 liters / hour and model boiler specimen MH-542-4 with a bobbin voltage of 503 volts which leaked at a rate of 0.86 liters / hour. These values can be compared to the Farley Unit 1 pulled tube with a voltage of 13.7 volts and a leak rate of 164 liters (0.72 gpm) at steam line break differential pressures.
The low leakage rates coupled with the high bobbin voltages result in the inability to satisfy the Generic Letter 95-05 statistical criteria. If the two data points had not leaked at all or had leaked at significantly higher leak rates, the current issue would not exist.
The projections for leakage in the event of a steam line break are 95%/95% projections of the steam line break leakage. Consequently, it is anticipated that the primary-to-secondary leakage should a steam line break occur would be less that that projected.
The limiting calculation for determination of the steam line break leakage limit for Farley is the accident initiated iodine spike. The accident initiated iodine spike limit is based on not exceeding a small fraction, i.e.,10%, of the 10 CFR 100 dose limits.
O Enclosure 2 l
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i Basis for Amendment i
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Basis for Amendment Introduction implementation of the voltage based repair criteria for Westinghouse steam generators affected by outside diameter stress corrosion cracking rcquires the implementation of a primary-to-secondary leakage limit in the event of a steam line break outside containment. By limiting the primary to secondary leakage, the radiological consequences of a postulated accident will not execcd the guideline values of 10CFR100. The quantity of primary to secondary leakage that is acceptable in the event of a postulated accident is dependent on the allowable activity in the primary coolant.
Descrintion of the Amendment Reauest
%c proposed amendment will modify Technical Specification 3/4.4.9, " Specific Activity," and the associated bases to reduce the limit associated with dose equivalent iodine 131. The steady state dose equivalent iodine 131 limit will be reduced to 0.15 prwic/ gram. The transient limit for 80% to 100%
power provided by Technical Specification Figure 3.4 1 will be reduced to 9 pCi/ gram. For less than 80%
power, the transient limits will also be appropriately reduced.
Halli By letter dated June 4,1992, Southern Nuclear submitted an assessment of the radiological dose consequences of a main steam line break. Wat calculation, based on a dose equivalent iodine 131 limit of 1.0 pCuric/ gram for steady state with a corresponding transient limit, concluded that the leak rate in the faulted steam generator should be limitod to 5.7 gpm.
By letter dated April 5,1994, the NRC Staffissued amendment 106 to the Farley Unit I technical specifications. This amendment reduced the dose equivalent iodine 131 limit to .25 pCuric/ gram for steady state with a corresponding reduction in the transient limit. This resulted in a leakage limit of(4 x 5.7 gpm) or 22.8 gpm.
By letter dated September 28,1995, the NRC Staffissued amendment 117 for the Farley Unit I technical specifications. This amendment increased the dose equivalent iodine 131 limit to .5 pCuric/ gram for steady state with a corresponding increase in the transient limit. This resulted i:' a leakage limit of(2 x 5.7 gpm) or 11.4 gpm. This safety evaluation inch.ded a description of confirmatory calculations perfomied by the NRC Staff that " concluded that a leak rate of i 1.4 gpm is an acceptable limit of the maximum primary to secondary leakage initiated by the steam line break accident."
By letter dated May 19,1997, the NRC Staffissued amendment 128 for the Farley Unit I technical specifications. This amendment reduced the dose equivalent iodine-131 limit to 0.3 pCuric/ gram for steady state with a corresponding decrease in the transient * .a. This resulted in a leakage limit of(3.333 x 5.7 gpm) or 19 gpm.
Similar technical specification changes have been approved for Farley Unit 2.
Subsequent to this amendment, it was discovered that steam line break leakage limit and steam line break leakage projections had not consistently accounted for the density of water at various temperatures. This results in a reduction of the steam line break leakage limit to value that is 71% of the 19 gpm value calculated above, or 13.5 gpm.
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his proposed amomiment will result in a slose equivalent iodine 131 limit of 0.15 pCuric/ gram for steady state and a corresponding reduction in the transient limit of 9 pCi/ gram. His will result in a density compensated leakage limit of 23.8 gpm. De transient limit will be 9 pCurie/ gram above 80% power with limits as shown on Figure 3.4 1 below 80% power. Ec calculation of these limits is consistent with that submitted originally by Southern Nuclear letter dated June 4,1992 and subsequently discussed in safety evaluations issued by the NRC Staffin letters dated April 5,1994 and September 28,1995.
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Enclosure 3 Significant flazards Evaluation v v
primary Coolant Specific Acthity Limit Reduction Farley Nuclear plant Significant flazards Evaluation Introduction implementation of the voltage based repair criteria for Westinghouse steam generators afTected by outside diameter stress corrosion cracking requires the implementation of a primary to-secondary leakage limit in the event of a steam line break outside containment. By limiting the prinary to sccondary leakage, the radiological consequences of a postulated accident will not exceed the guideline vahes of 10CFR100. He quantity of primary to-secondary leakage that is acceptable in the event of a postulatui accident is dependent on the allowable activity in the primary coolant.
Description of the Amendment Reauest As required by 10CFR50.91(a)(l), this analysis is provided to demonstrate that a proposed amendment to reduce the dose equivalent lodine limits for Farley Units I and 2 represents no significant hazards, in accordance with 10CFR50.92(c), implementation of the proposed license amendment was anal . zed using the following standards and found not to: 1) involve a significant increase in the probability or consequences for an accident previously evaluated; 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.
He proposed amendment will modify Technical Specification 3/4.4.9, " Specific Activity," and the t
associated bases to reduce the limit associated with dose equivalent iodine-131. The steady state dose equivalent iodine 131 limit will be reduced to 0.15 pCurie/ gram. The transient limit for 80% to 100%
power provided by Technical Specification Figure 3.4 1 will be reduced to 9 pCi/ gram. For less than 80%
power, the transient limits will also be appropriately reduced.
3 Evaluation By letter dated June 4,1992, Southern Nuclear submitted an assessment of the radiological dose consequences of a main steam line break. That calculation, based on a dose equivalent iodine-131 limit of 1.0 pCurie/ gram for steady state with a corresponding transient limit, concluded that the leak rate in the faulted steam generator should be limited to 5.7 gpm.
By letter dated April 5,1994, the NRC Staffissued amendment 106 to the Farley Unit I technical specifications, His amendment reduced the dose equivalent iodine 131 limit to .25 pCurie/ gram for steady state with a corresponding reduction in the transient limit. This resulted in a leakage limit of(4 x 5.7 gpm) or 22.8 gpm.
By letter dated September 28,1995, the NRC StafTissued amendment i 17 for the Farley Unit I technical specifications. This amendment increased the dose equivalent iodine-131 limit to .5 pCurie/ gram for steady state with a corresponding increase in the transient limit. This resulted in a leakage limit of(2 x 5.7 gpm) or 11.4 gpm. This safety evaluation included a description of confirmatory calculations performed by the NRC Staff that " concluded that a leak rate of i 1.4 gpm is an acceptable limit of the maximum primary to secondary leakage initiated by the steam line break accident."
By letter dated May 19,1997, the NRC StafTissued amendment 128 for the Farley Unit I technical specifications. This amendment reduced the dose equivalent iodine-131 limit to 0.3 pCurie/ gram for steady state with a corresponding decrease in the transient limit. This resulted in a leakage limit of(3.333 x 5.7 gpm) or 19 gpm.
Significant llazards Evaluction page 2 Raluction in Dose Equivalent lodine 131 Limits Similar tecimical specification changes have been approved for Farley Unit 2.
Subsequent to this amendment, it was discovered that steam line break leakage limit and steam li,e break leakage projections had not consistently accounted for the density of water at various temperatmes. His results in a reduction of the steam line break leakage limit to value that is 71% of the 19 gpm value calculated above, or 13.5 gpm.
His proposed amendment will result in a dose equivalent iodine 131 limit of 0.15 pCuric/ gram for steady state and a corrcsponding reduction in the transient limit. His will result in a density compensated leakage limit of 23.8 gpm. %c transient limit will be 9 pCuric/ gram above 80% power with limits as shown on Figure 3.4 1 below 80% power. He calculation of these limits is consistent with that submitted originally by Southern Nuclear letter dated June 4,1992 and subsequently discussed in safety evaluations issued by the NRC Staffin letters dated April 5,1994 and September 28,1995.
Analysis Conformance of the proposed amendment to the standards for a detennining that it involves no significant hazards a: defmed in 10CFR50.92 is shown by the following.
- 1. Operation of Farley Units I and 2 in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
He reduction in the dose equivalent iodine limits, both steady state and transient, will not increase the probability of any accident evaluated since no physical changes to the plant are being made. He consequences of any accident previously evaluated will not be increased since the speci6c activity limit of the primary coolant is being decreased.
- 2. He proposed license amendment does not create the possibility of a new or different kind of accident from any accident presiously evaluated.
He reduction in the dose equivalent iodine limits, both steady state and transient, will not create the possibility of a new or difTerent kind of accident from any accident previously evaluated since no physical changes to the plant are being made. He accidents of concem continue to be those that have previously been analyzed.
- 3. The proposed license amendment does not involve a signi6 cant reduction in a margin of safety.
He calculated potential radiological consequences from the main steam line break tecident remain within the regulatory exposure guidelines and have not changed. Reduction of the dose equimlent iodine limit to increase allowable steam line break primary to secondary steam generator leakage is a compensating offsite dose effect. Consequently, there is no reduction in any margin of safety.
Conclusion Based on the preceding analysis, it is concluded that operation of Farley Units I and 2 following the reduction in the dose equivalent iodine 131 limits in accordance with the proposed amendment does not increase the probability of an accident previously evaluated, create the possibility of a new or different kind ofacet' ' 1m any accident presiously evaluated, or reduce any margins of safety. Acrefore, the license amend' does not involve any significant hazards as defmed in 10CFR50.92.