View source for L-99-018, Application for Amend to License NPF-2 for Approval to Operate Unit 1,cycle 16 Only,Based on risk-informed Approach for Evaluation of SG Tube Structural Integrity

{{Adams
| number = ML20206G284
| issue date = 04/30/1999
| title = Application for Amend to License NPF-2 for Approval to Operate Unit 1,cycle 16 Only,Based on risk-informed Approach for Evaluation of SG Tube Structural Integrity
| author name = Morey D
| author affiliation = SOUTHERN NUCLEAR OPERATING CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000348
| license number = 
| contact person = 
| document report number = NEL-99-0183, NEL-99-183, NUDOCS 9905070152
| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS &amp; PERMITS
| page count = 12
| project = 
| stage = Request
}}

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9 Dav) Mor:y                   SIuthern Nuclear                                                           i Vice President               Operating Company,Inc.                                                   )
Farley Project               Post Office Box 1295                                                      l Birmingham, Alabama 35201 Tel 205.992.5131 i
SOUTHERN April 30, 1999-COMP M Energy to Serve YourWorld" Docket No.:              '50-348'                                                  NEle99-0183                  l U. S. Nuclear Regulatory Commission A'1TN: Document Control Desk Washington, DC 20555.
Joseph M. Farley Nuclear Plant - Unit 1 License Amendment Aeolication for Cvele 16 Ladies and Gentlemen:
In accordance with the provisions of 10 CFR 50.90, Southern Nuclear Operating Company (SNC), as the licensed operator for the Joseph M. Farley Nuclear Plant (FNP), hereby submits the enclosed application for amendment of the Facility Operating License No. NPF-2 for FNP Unit 1.
l This application requests Nuclear Regulatory Commission approval for SNC to operate FNP Unit 1, for cycle 16 only, based on the risk-informe/. approach for the evaluation of steam generator (SG) tube structural integrity as described by NEI 97-06, " Steam Generator Program Guidelines."
l The risk-informed probability information provided in this submittal, and the cycle length                       J evaluation analysis, submitted by SNC to the NRC on April 23,1999, provMe the technical basis
                     - for this submittal. The proposed amendment would constitute an additional condition being added to the FNP, Unit I license which would allow cycle 16 operation based on the risk-informed probability of tube rupture and nominal accident-induced primary-to-secondary leakage in the event of a steam line break.
Enclosure 1 contains the basis for the proposed license amendment. ' The supporting significant hazards evaluation pursuant to 10 CFR 50.91 i' povided in Enclosure 2. Based upon the analysis provided, SNC has determined that the proposed changes to the Facility Operating License No.
NPF-2 do not involve a significant hazards consideration as defined by 10 CFR 50.92. SNC has determined that the proposed license amendment will not significantly affect the quality of the
                     - human environment.
The proposed amended license typed page is included in Enclosure 3. The rrarked page of the current license, which denotes the proposed amendment, is provided in Enclosure 4.
The Plant Operations Review Committee has reviewed and recommended apprnval of these proposed changes. A copy of these proposed changes is being sent to Dr. Donald E. Williamson, the Alabama State Designee, in accordance with 10 CFR 50.91(b)(1).
                     ~ As discussed with the starTduring the April 16,1999 meeting, < pproval of this license amendment is requested by July 31,1999.
                                                                           ,,,n                                                    I 9905070152 990430                                                                                            Of PDR      ADOCK 05000348                                                                                              (
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         .,   . p. S. Nuclear Regulatory Commission if there are any questions, please advise.
Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY Dave Morey EWC/mafcyc16. doc Sworn to andsubscribedbefore me this & day of&          1999 WAA No~taly Public V
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My Commission Expires: f0' W $* L._               /, kOl Enclosu.w:
1.~ Basis for Proposed License Amendment
: 2. 10 CFR 50.92 Evaluation
: 3. . Proposed Amended License Typed Page
: 4. Marked Page of the Current License cc:     See next page, l
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            -Page 3
          . JJ. S. Nuclear Regulatory Commission cc:    Southern Nuclear Operatina Comoany Mr. L. M. Stinson, General Manager - Farley LLSyuclear Renulatory Commission. Washinnton. D. C.
Mr. J. J. Zimmerman, Licensing Project Manager - Farley U. S. Nuclear Regulatory Commission. Renion II Mr. L. A. Reyes, Regional Administrator Mr. T. P. Johnson, Senior Resident Inspector - Farley Alabama Department of Public Health Dr. D. E. Williamson, State Health Officer i
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ENCLOSUREI l
            .,                                   Joseph M. Farley Nuclear Plant License Amendment Application for Cycle 16 Basis for Pronosed License Amendment By {{letter dated|date=December 22, 1998|text=letter dated December 22,1998}}, which transmitted LER 98-007- 00, Southern Nuclear Operatmg Company (SNC) stated that the actual operating time until the next steam generator inspection would be based on the conclusions of the fmal condition monitoring and operating assessment evaluation. De results of the final Unit 1 Cycle 16 Freespan ODSCC Operational Assessment (OA) report and a severe accident risk evaluation performed to complement the OA were discussed during a meeting with the NRC Staff on April 16,1999 and submitted to the NRC on April 23,1999. Hat letter stated that it is SNC's perspective that operation for a full cycle for Farley Nuclear Plant (FNP) Unit 1, Cycle 16, is within the current licensing basis. However, SNC recognizcs that the NRC Staff has expressed uncertainty with this opinion.
To address NRC Stafflicensing basis concerns, this proposed Facility Operating License amendment for         j the current operating cycle is requested. The change will allow for a risk-informed approach to the evaluation of steam generator tube structural integrity as described by NEI 97-06, " Steam Generator Program Guidelines." The risk informed probability information provided in this submittal, and the cycle length evaluation analysis, submitted by SNC to the NRC on April 23,1999, provide the technical basis for this submittal.
De changes proposed to the Facility Operating License reflect consideration for the probability of tube burst and accident induced tube leakage as a result of a steam line break. He probability of burst limit of  !
5.0 E-2 is based on NEI 97-06 guidelines for all steam generator tube degradation mechanisms. He calculated probability of burst for freespan flaws, when added to the probability of burst for ODSCC at tube support plate intersections, is also sufficiently small to provide adequate margin to the DG-1074 proposed limit of 2.5 E-2 to account for corrosion degradation. The accident induced leakage limit of a nominal 1 gpm provides assurance that the dose rate at the site boundary will not exceed the guidance of
      .10 CFR 100. Sufficient margin exists such that the sum of the freespan projected leak rate and the projected leak rate for ODSCC at the tube support plates will remain less than the recently reduced 11.8 gpm licensing based limit.
l
      - As proposed in Draft Regulatory Guide DG 1074, " Steam Generator Tube Integrity," SNC developed a severe accident risk assessment to complement our probabilistic steam generator tube operational assessment.' The assessment considered accident sequences that are contributors to the core damage frequency (CDF) currently assumed in the FNP probabilistic risk assessment (PRA). Since many of the sequences are beyond FSAR Chapter 15 accident analyses assumptions (current licensing basis), they are given the term severe accidents. The methods used to perform this assessment are similar to the approach    j used in NUREG 1570," Risk Assessment of Severe Accident - Induced Steam Generator Tube Rupture."            l Insights gained through the study of NUREG 1570 assisted in focusing this risk assessment to those areas    l of greatest risk. For FNP, considering severe accident containment bypass due to High/ Dry (H/D) conditions and the use of a probability of tube burst approach to determine SG structural integrity, the total Large Early Release Fraction (LERF) is 2.68 E-6/ Reactor Year (Rx-Yr). Using the guidance found in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," this change is considered to be       ,
small and in Region II of Figure 4.0. He Region II acceptance criteria is defmed by calculated increases    l El-1 4

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;.         Joseph M. Farley Nuclear Plant
!          License Amendment Application for Cycle 16 Basis for Prooosed License Amendment l
in LERF in the range of E-7 per reactor year to E-6 per reactor year. RG 1.174 also states for Region II that the total LERF is less than E-5 per reactor year. Therefore, based on the results of this evaluation,     l use of a probability of tube burst approach to determine steam generator structural integrity for FNP Unit 1 end of cycle 16 is appropriate and results in only a small change in LERF. No changes to CDF occurred as a result of this analysis.
The results of the steam generator operational assessments indicate that operation for a full operating cycle resulu in acceptable tube burst probabilities and that the severe accident risk is small and within the acceptance criteria of RG 1.174. These results ensure that the changes proposed in this amendment are acceptable.
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m ENCLOSURE 2
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Joseph M. Farley Nuclear Plant License Amendment Application for Cycle 16 10 CFR 50.92 Evaluation Puisuant to 10 CFR 50.92 cach application for amendment to an operating license must be reviewed to determine if the proposed change involves a significant hazards consideration. The amendment, as defined below, describing the proposed amendment to the Facility Operatmg License NPF-2, has been reviewed and deemed not to involve a significant hazards consideration. De basis for this determination follows.
BACKGROUND By {{letter dated|date=December 22, 1998|text=letter dated December 22,1998}}, which transmitted LER 98-007- 00, Southern Nuclear Operating Company (SNC) stated that the actual operating time until the next SG inspection would be based on the conclusions of the final condition monitoring and operating assessment evaluation. The results of the      ,
final Unit 1 Cycle 16 Frecspan ODSCC Operational Assessment (OA) report and a severe accident risk         I evaluation performed to complement the OA were discussed during a meeting with the NRC Staff on             l April 16,1999 and submitted to the NRC on April 23,1999. That letter stated that it is SNC's               !
perspective that operation for a full cycle for Farley Nuclear Plant (FNP) Unit 1, Cycle 16, is within the  I current licensing basis. However, SNC recognizes that the NRC Staff has expressed uncertainty with this opinion.
To address NRC licensing basis concerns, this proposed Facility Operating License amendment for the current operating cycle is requested. The change will allow for a risk-informed approach to the evaluation of steam generator tube stmetural integrity as described by NEl 97-06, " Steam Generator Program Guidelines." He risk-informed probability information provided in this submittal, and the cycle -
length evahaation analysis, submitted by SNC to the NRC on April 23,1999, provide the technical basis for this submittal.
DESCRIPTION OF CHANGE REOUEST The proposed amendment to Facility Operating License NPF-2 for FNP constitutes an additional condition being added,2.C.(3)(i), which states "For cycle 16 only, SNC shall be permitted to operate the reactor based on the risk-informed probability of steam generator tube rupture for all mechanisms of s 5.0 E-2 and nominal freespan accident-induced primary-to-secondary leakage of s 1 gallon per minute in the event of a steam line break." Additionally, to avoid confusion with the acceptable operating leakage limit, the condition explicitly states that 'The accident-induced, primary-to-secondary leakage would be in addition to the operational leakage limits that are stipulated in the Technical Specifications."
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Joseph M. Farley Nuclear Plant License Amendment Application for Cycle 16
    .,     10 CFR 50.92 Evaluation ANALYSIS l
He FNP Technical Specifications provide that a 40% through wall tube indication repair criterion is           l acceptable for the FNP steam generators. The NRC Staff position is that 3 times normal operating             j differential pressure across the SG tubes (3AP) is the basis for d.c 40% repair criterion. SNC performed     j an analysis that projected the probability of meeting 3 AP at the end of the planned operating cycle. The analysis indicated that there is an 80% probability that the worst steam generator would meet the 3AP requirement at the end of the planned operating cycle. To provide further assurance, SNC perfvimcd an        s
       ' analysis to address the issues that the 3AP margin has been used to address (burst in the event of a steam    I line break and severe accident concerns).
Burst in the event'of a steam line break SNC performed an analysis to project the probability of burst in the event of a steam line break on the last day of the planned operating cycle. De probability of burst (95% probability at a 95% confidence level) is 1.1E-2. His results in a margin for unknown and unanalyzed causes of tube rupture of 3.4E-2. His result is within the guidelines proposed by NEI 97-06, " Steam Generator Program Guidelines."
Severe accident concerns                                                                                     ,
SNC performed an analysis to estimate the probability of containment bypass based on the plug-on-            l detection repair criteria used at the last inspection. The calculated probability of containment bypass change is less than IE4/Rx-Yr and total containment bypass is less than IE,5/Rx-Yr. His result is acceptable based on the guidance of Regulatory Guide 1.174,"An Approach for Using Probabilistic Risk         j Assessment in Risk-Informed Decision on Plant-Specific Changes to the Licensing Basis."                       ,
1 enm nominal orimary-to-secondary leakane he 1 gpm nominal primary-to-secondary leakage assumption is used in several design basis analyses contained in the FSAR. Consequently, SNC performed an analysis that projects primary-to-secondary leakage in the event of a steam line break on the last day of the planned operating cycle. Hat projected freespan ODSCC leak rate (95% probability at a 95% confidence) is less than 0.01 gpm. He projected end of cycle leakage for alternate repair criteria at tube support plate (ARC TSP) for Unit 1 is '8.2 gpm.
Total leakage from both freespan ODSCC and ARC TSP in the event of a mainsteam line break is calculated to be 8.21 gpm. This value is well below the leakage value that the NRC Staff has approved for ARC TSP, which is based on offsite dose calculations not exceeding a small fraction of 10 CFR Part 100 dose limits. A technical specification amendment was recently submitted to the NRC Staff to reduce the design basis limit from 23.8 to 11.8 gpm. Even with the lower limit sufficient margin exists to assure that offsite dose remains within a small fraction of 10 CFR Part 100 dose limits.
10 CFR 50.92 EVALUATION CONCLUSIONS Based on the preceding evaluation, the following conclusions are provided with respect to the criteria contained in 10 CFR 50.92.
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Joseph M. Farley Nuclear Plant License Amendment Application for Cycle 16
    .,   10 CFR 50.92 Evaluation The proposed changes do not significantly increase the probability or consequences of an accident previously evaluated in the FSAR. The probability of tube burst is slightly increased as a result of this proposed amendment but is within current industry guidance. Therefore, the probability of a previously evaluated accident are not significantly increased. There is no change in the FNP design basis as a result of this change and,~ as a result, this change does not involve a significant increase in the consequences of an accident previously evaluated.
De proposed changes to the TSs do not increase the possibility of a new or different kind of accident than any accident already evaluated in the FSAR. No new limiting single failure or accident scenario has been created or identified due to the proposed changes. Safety-related systems will continue to perform as designed. He proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.
He proposed changes do not involve a significant reduction in the margin of safety. %ere is no impact in the accident analyses. Hese proposed changes are technically consistent with the requirements of NEI 97-06, " Steam Generator Program Guidelines," Draft Regulatory Guide DG 1074, " Steam Generator Tube Integrity," and Regulatory Guide (RG) 1.174,"An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis." nus the proposed changes do not involve a significant reduction in the margin of safety; Accordingly, SNC has determined that the proposed amendment to the Facility Operating License NPF-2 does not involve a significant hazards consideration.
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ENCLOSURE 3                   l Joseph M. Farley Nuclear Plant License Amendment Application for Cycle 16 Proposed Amended License Tvoed Pane l
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y l
: 2. Identification of the procedures used to quantify parameters j
that are critical to control points;
: 3. Identification of process sampling points; 4      A procedure for the recording and management of data;
: 5. Procedures defining corrective actions for off control point chemistry conditions; and                                     .
: 6. A procedure ider.tifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events required to initiate corrective action.
(h)     The Additional Conditions contained in Appendix C, as revised through Amendment No.137, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the additional conditions.
1 (i)     For Cycle 16 only, Southern Nuclear shall be permitted to operate the reactor based on the risk-informed probability of Steam Generator tube rupture for all mechanisms ofs 5.0 E-2 and nominal freespan accident-induced primary-to-secondary leakage ofs 1 gallon per minute in the event of a steam line break. The accident-induced, primary-to-secondary leakage limit would be in adi ' to the ce erational leakage limits that are stipulated in the Technical Specific vions.
(4) Fire Protection Southern Nuclear shall implement and maintain in effect all provisions of the approved fire protection program as described on the Final Safety Analysis Report for the facility and as approved in the Fire Protection Safety Evaluation Reports dated February 12,1979, August 24,1983, December 30,1983, November 19,1985, September 10,1986, and December 29,1986. Southern Nuclear may make changes to the approved             i fire protection program without prior approval of the Commission only if        !
those changes would not adversely affect the ability to achieve and             l maintain safe shutdown.                                                         l l
Farley - Unit 1                                                           Amendment No.

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ENCLOSURE 4 l
Joseph M.Farley Nuclear Plant License Amendment Application for Cycle 16 Marked Pane of the Current License l
l a
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: 2.      Identification of the procedures used to quantify parameters that are critical to control points;
: 3. Identification of process sampling points;
: 4. A procedure for the recording and management of data;
: 5. Procedures defining corrective actions for off control point chemistry conditions; and
: 6. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events required to initiate corrective action.
(h)     The Additional Conditions contained in Appendix C, as revised through Amendment No.137, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the additional conditions.
(i)     For Cycle 16 only, Southern Nuclear shall be permitted to operate the reactor based on the risk-informed probability of Steam INSERT                  Generator tube rupture for all mechanisms ofs 5.0 E-2 and
                           >            nominal freespan accident-induced primary-to-secondary leakage ofs 1 gallon per minute in the event of a steam line break. The accident-induced, primary-to-secondary leakage limit would be in addition to the operational leakage limits that are stipulated in the
(          Technical Specifications.                                             j (4)     Fire Protection Southern Nuclear shall implement and maintain in effect all provisions of the approved fire protection program as described on the Final Safety Analysis Report for the facility and as approved in the Fire Protection Safety Evaluation Reports dated February 12,1979, August 24,1983,
                             . December 30,1983, November 19,1985, September 10,1986, and December 29,1986. Southern Nuclear may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown.
Farley- Unit 1                                                               Amendment No.141 l}}