View source for NOC-AE-000097, Application for Amends to Licenses NPF-76 & NPF-80, Incorporating Tech Specs 3.4.5 Re 1-volt voltage-based Repair Criteria for SG Tube Support plate-to-tube Intersections,Per GL 95-05

{{Adams
| number = ML20217P115
| issue date = 04/02/1998
| title = Application for Amends to Licenses NPF-76 & NPF-80, Incorporating Tech Specs 3.4.5 Re 1-volt voltage-based Repair Criteria for SG Tube Support plate-to-tube Intersections,Per GL 95-05
| author name = Cloninger T
| author affiliation = HOUSTON LIGHTING &amp; POWER CO.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000498, 05000499
| license number = 
| contact person = 
| document report number = GL-95-05, GL-95-5, NOC-AE-000097, NOC-AE-97, NUDOCS 9804090324
| package number = ML20217P123
| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS &amp; PERMITS
| page count = 22
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:l               l                                                                                                   1
              .J Nuclear Operaung Company South Taasnr$t1Ektet GemrathyStatkm 19 Bar239 Kaakarth. Taxs77483                             m i
April 2,1998 NOC-AE-000097 File No.: G20.02.01 G20.02.02 10 CFR 50.90 l
U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington,DC 20555 South Texas Project Units I and 2 Docket Nos. STN 50-498, STN 50-499 Proposed Amendment to Incorporate Voltage-Based Repair Criteria into Technical Snecification 3.4.5 South Texas Project (STP) proposes to amend the Technical Specifications by incorporating the attached amendment to Technical Specification 3.4.5. These changes will implement 1-volt voltage-based repair criteria for the steam generator tube support plate-to-tube intersections for Unit 2 and make related Unit I administrative changes for consistency of wording. The proposed amendment is consistent with guidance provided in Generic Letter 95-05.
The attached application supplements the Technical Specification amendment proposed by STP in {{letter dated|date=February 16, 1998|text=letter dated February 16,1998}} (NOC-AE-0074). This supplement primarily clarifies that the
; changes proposed by the February 16,1998 submittal are technically applicable to both Units l
and provides additionaljustification for a number of the changes. To assist your review, the changes made to the amendment originally submitted on February 16,1998, are indicated by revision bars. This supplement makes no revisions to the Technical Specifications and Bases proposed in the February 16,1998 submittal.
STP has reviewed the proposed amendment pursuant to 10CFR50.92 and determined that it does not involve a significant hazards consideration. In addition, STP has determined that the proposed amendment satisfies the criteria of 10CFR51.22(c)(9) for categorical exclusion from the requirement for an environmental assessment. The STP Plant Operations Review Committee and the Nuclear Safety Review Board have reviewed and approved the proposed amendment.
The required affidavit, a Safety Evaluation and Determination of No Significant Hazards Consideration, and the marked-up affected pages of the Technical Specifications are included as
                                                                                                           \g         '
f    g'!
attachments to this letter. Also attached is Westinghouse Steam Generator Report SG-98-01-004                o~
a., V   J g\
                                       ,4 9804090324 980402                                                                        sTI. 30547247 PDR      ADOCK 05000498 P                           PDR I

I l                                                                                          NOC-AE-000097 Page 2 of 4
!                                                                                                              i j
which provides the technical basis supporting the application of Generic Letter 95-05 to STP
     . Unit 2. Marked-up pages from both the current Technical Specifications and the Improved                 !
l     Technical Specifications have been attached to reflect the proposed changes, although l
l
    . Attachment 5 is included for information only.                                                           l 1
,     In accordance with 10 CFR 50.91(b), STP is notifying the State of Texas of this request for license
!-    amendment by providing a copy of this letter and its attachments.
l     STP requests that this proposed amendment be reviewed and approved by June 1,1998 to allow               1 i
procedures, non-destructive examination (NDE) analysis guidelines, and the analyst training              !
program to be revised to reflect the voltage-based repair criteria for the upcoming Unit 2               :
: refueling outage (2RE06). STP also requests 30 days for implettentation.
:l If there are any questions regarding the proposed amendment, please contact Mr. M. A. McBurnett at (512) 972-7206 or myself at (512) 972-8787.
Y                    .
l
                                                                      . H. Cl inger Vice P side       ucle Engineering jtc/
Attachments: 1. Description of Amendment Request                                         ,
: 2. Safety Evaluation
: 3. Determination of No Significant Hazards Consideration
: 4. Proposed Changes to Current Technical Specifications
: 5. Proposed Changes to Improved Technical Specifications
: 6. Westinghouse Steam Generator Report SG-98-01-004 l
E:\WP\NL\NRC-WK\TsC-98\0097. DOC                                                           STI: 30547247

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i NOC-AE-000097 Page 3 of 4 o
cc:
Ellis W. Merschoff '                        Jon C. Wood
     . Regional Administrator, Region IV'          Matthews & Branscomb
;      U. S. Nuclear Regulatory Commission         One Alamo Center
{
611 Ryan Plaza Drive, Suite 400              106 S. St. Mary's Street, Suite 700
      - Arlington, TX _76011-8064                  San Antonio, TX 78205-3692 Thomas W. Alexion-                          Institute of Nuclear Power Operations           ;
,      Project Manager, Mail Code 13H3             Records Center ~                                j U. S. Nuclear Regulatory Commission -       700 Galleria Parkway-                          1 Washington, DC 20555-0001                   Atlanta, GA 30339-5957
                                                                                                   )
!                                                                                                  i David P. Loveless                           Richard A. Ratliff Senio'r Resident Inspector                  Bureau of Radiation Control U. S. Nuclear Regulatory Commission         Texas Department of Health P. O. Box 910
{
1100 West 49th Street Bay City, TX 77404-0910                    . Austin, TX ~ 78756-3189                        !
J._R. Newman, Esquire                       D. G. Tees /R. L. Balcom Morgan, Lewis A.Bockius                     Houston Lighting & Power Co.'                   l 1800 M Street, N.W.                         P. O. Box 1700                                  i Washington, DC 20036-5869                   Houston,TX 77251                                !
I M. T. Hardt/W. C. Gunst                     Central Power and Light Company                 ;
City Public Service                         Attn: G. E. Vaughn/C. A. Johnson                !
P. O. Box 1771                              P. O. Box 289, Mail Code: N5012                 i San Antonio,TX 78296 ~                      Wadsworth,TX 77483 l
J. C. Lanier/A. Ramirez                      U. S. Nuclear Regulatory Commission City of Austin Electric Utility Department  Attention: Document Control Desk 721 Barton Springs Road                     Washington, DC '20555-0001                      l
     ' Austin, TX 78704                                                                            !
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4
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l E;\WPWL\NRC-WK\TsC.98\0097. doc                                             sTI: 30547247 I-

c NOC-AE-000097   ;
I Page 4 of 4 l
l                                                UNITED STATES OF AMERICA l                                        NUCLEAR REGULATORY COMMISSION l
L In the Matter of                                            )
                                                              )
STP Nuclear Operating Company                               )                   Docket Nos. 50-498            l
                                                              )                                 50-499          {'
South Texas Project Units 1 & 2                             )
l                                                        AFFIDAVIT t
I, T. H. Cloninger, being duly sworn, hereby depose and say that I am Vice President, Nuclear Engineering, of STP Nuclear Operating Company; that I am duly authorized to sign and file with the Nuclear Regulatory Commission the attached proposed amendment to the Technical l Specifications; that I am familiar with the content thereof; and that the matters set forth therein are true and correct to the best of my knowledge and belief.
In                                    '
l
                                                                . Idioliing ice Presid   t, Nucl
                                                                                           }
Engi6eering             i I
l l
l STATE OF TEXAS                                     )
                                                     )
COUNTY OF MATAGORDA                                )
Subscribeg and swom to before me, a Notary Public in and for the State of Texas, this 2 day of f(pn /   ,
                                      ,1998.
          / kI' UNDARITTENBERRY
       ',           , Notary Pute:, State of Texas                  d
            ,    ,p Q{[                                Notary Public in and for the f State of Texas EAWPNL\NRC WK\TsC-98\0097. DOC                                                                  sTl: 30547247

ATTACHMENT 1 1
l DESCRIPTION OF AMENDMENT                      I
'                                                          4 REQUEST I
i 1
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E:\WP\NL\NRC-WK\TSC-98WO97. DOC            STI: 30547247

[                                                                                                            I l
!i                                                                                          NOC-AE-000097 Attachment 1 Page1of5 DESCRIPTION OF AMENDMENT REQUEST The proposed amendment would revise Technical Specifications 3/4.4.5 and its Basis to allow the implementation of 1-volt voltage-based repair criteria for the steam generator tube support plate-to-tube intersections for Unit 2 in accordance with Generic Letter 95-05. The proposed amendment also includes an administrative change to Basis 3.4.6.2 to clarify that the allowable steam generator leakage specification applies to both units at STP. This amendment is consistent with the guidance provided in NRC Generic Letter 95-05, " Voltage-Based Repair Criteria for l   Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Conosion l  Cracking." As discussed below. the only pertinence to Unit 1 are:
l
!       e   minor wording changes (e.g.. removal of"For Unit 1"1 e   clarification in Section 4.4.5.4.a.12 that the plugging limit is used for mill annealed allov 600 tubes.
l
        . two non-technical changes discussed below. and e   the use of an alternative POD (probability of detection of ODSCC flawst i
l l                                            BACKGROUND Previous inservice inspections and examinations of the steam generator tubes have identified intergranular stress corrosion cracking (IGSCC) on the outer diameter of the tubes at the tube support plate (TSP) intersections. This particular form ofIGSCC is known as outer diameter stress corrosion cracking (ODSCC) and is a degradation phenomenon found in a number of                    !
nuclear power plant steam generators. Various tubes, including tube-to-TSP intersections, have i
been removed from affected steam generators in numerous nuclear plants for examination and l  testing. Each of the pulled tubes was sectioned and metallographically examined. The
;  examinations have revealed multiple, segmented, and axial cracks with short lengths for the l  deepest penetrations. The ODSCC is generally confined within the thickness of the TSPs, I
consistent with the corrosion mechanism which involves the concentration ofimpurities, including caustics, in the tube-to-TSP crevices. There is some potential for shallow ODSCC for l  a short distance above or below the TSP. This has been observed in the TSP intersections of some pulled tubes from another plant.                                                                     I
;  The steam generator tube specimens pulled from STP Unit 1 in 1993 and 1995 have shown only                l l
limited intergranular attack (IGA) associated with the ODSCC. However, more significant IGA has been observed to occur occasionally with ODSCC on some pulled tube specimens from other l  plants. These results suggest that in some cases the degradation developed as IGA plus stress             ;
j corrosion cracking (SCC). This combination ofIGA plus SCC was seen when maximum IGA l  depths were greater than 25 percent. A large number (> 100) of axial cracks around the circumference are commonly found on these tubes. The maximum depth ofIGA is typically one-half to one-third of the SCC depth. Patches of cellular IGA /ODSCC formed by combined axial and circumferential orientation of microcracks are occasionally found in pulled-tube                I examinations. Axial crack segments have been the dominant flaw feature affecting the structural E:\WMNL\NRC-WK\TsC-98\0097. doc                                                             sTI: 30547247

NOC-AE-000097 Attachment 1 Page 2 of 5 integrity of the pulled-tube specimens as evidenced by results of burst tests of the pulled TSP intersections prior to sectioning. Testing of tubes with ODSCC has demonstrated a high margin to failure and evaluations have shown that existing tube plugging criteria would cause unnecessary and inappropriate tube plugging.
DESCRIPTION OF CIIANGES The proposed amendment modifies the steam generator surveillance requirements to allow implementation of the tube support plate voltage-based repair criteria for Unit 2. These surveillance requirements have been previously approved for Unit 1.
The 100% bobbin coil inspection for all hot leg tube support plate intersections and all cold leg intersections down to the lowest cold leg tube support plate with known ODSCC indications in Surveillance Requirement 4.4.5.2.e now applies to both units.
The definitions of plugging limit and imperfection depths in Surveillance Requirement 4.4.5.4.a.7 now apply to both units.
The Tube Support Plate Plugging Limit described in Surveillance Requirement 4.4.5.4.a.12 now applies to both units. Note I has been amended to remove tube diameter information not applicable to STP that was introduced by Generic Letter 95-05 sample Technical Specifications.
Section 4.4.5.4.a.12 clarifies that the pluggine limit is applicable to mill annealed alloy 600 tubes. as noted in item 2 in the Justification and the topical report associated with this application.
A grammatical error is corrected with the addition of the word "of"in Section 4.4.5.4.a.12(c).
Section 4.4.5.4.a.12(d) reflects that certain intersections in Unit 1 are excluded from application of the voltage-based repair criteria.
The additional reporting criteria in Surveillance Requirement 4.4.5.5.d now apply to both units.
The Bases for 3/4.4.5 and 3/4.4.6.2 reflect that voltage-based repair criteria and steam generator tube leakage limits now apply to both units.
E:\WPNL\NRC-WK\TsC-98WO97. doc                                                             sTI: 30547247

i            ,
NOC-AE-000097 Attachment 1 Page 3 of 5 l
JUSTIFICATION Technical Specification 4.4.5.4.a.7, Plugging Limit, requires that tubes with imperfections exceeding 40% of the nominal tube wall thickness be removed from service. This criterion would result in unnecessarily plugging significant numbers of steam generator tubes affected with ODSCC at TSPs. Unnecessarily plugged tubes reduce steam generator heat removal capability in both accident conditions and during normal operations. To preclude this reduced
                                                                                                                      ^
:        capability, STP proposes voltage-based repair criteria for Westinghouse steam generator tubes l        affected by ODSCC.
L,       Voltage-based repair criteria for Westinghouse steam generator tubes affected by ODSCC                     ,
l        involves a correlation between eddy current bobbin coil signal amplitude (voltage) versus tube i        burst pressure and leak rate. The principal parameter is bobbin voltage amplitude.which is L     ' correlated with tube burst capability and leakage potential. The voltage-based repair criteria are developed by EPRI from testing oflaboratory-induced ODSCC specimens and extensive examination ofpulled tubes from operating steam generators.
The voltage-based repair criteria are based on compliance with the NRC Generic Letter 95-05 and are described in the attached Westinghouse Steam Generator Report SG-98-01-004. The methodology employed follows the industry degradation-specific management methodology developed by EPRI and is similar to that implemented for nine plants, four of which had 3/4" diameter steam generator tubing. The bobbin coil voltage criteria detailed in this proposed l
amendment to the Technical Specifications reflects a conservative approach for the STP voltage-based repair criteria, recognizing that higher limits have been demonstrated to provide adequate margins in accordance with applicable regulatory requirements. The proposed voltage-based repair criteria f= Ud 2 are provided in accordance with the following:                                   l
: 1.      Implementation of the steam generator tube / tube support voltage-based repair criteria L                requires a 100% bobbin coil inspection for all hot leg tube support plate intersections and l                all cold leg tube support plate intersections down to the lowest cold leg tube support plate j-               with ODSCC indications. The determination of the lowest cold leg tube support plate intersections having ODSCC indications shall be based on the performance of at least a
                - 20% random sampling of tubes inspected over their full length.
L      2.        The tube support plate voltage-based repair criteria limit is used for the disposition of a
!                mill annealed alloy 600 steam generator tube for continued service that is experiencing                 ,
L                predominately axially oriented ODSCC confined within the thickness of the tube support -
l   3            plates. At tube support plate intersections, the plugging (repair) limit is based on maintaining steam generator tube serviceability as described below:                                    '
a)       Steam generator tubes whose degradation is attributed to ODSCC within the bounds of the tube support plate with bobbin voltage less than or equal to the lower voltage repair limit (1 volt) will be allowed to remain in service.
EAWP(NL\NRC-WK\Tsc-98iOO97. doc                                                              sTI: 30547247

NOC-AE-000097 i                                                                                         Attachment 1 Page 4 of 5 b)        Steam generator tubes whose degradation is attributed to ODSCC within the
                   . bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit will be repaired or plugged except as noted in item 2.c below, c)        Steam generator tubes, with indications of potential degradation attributed to          l ODSCC within the bounds of the tube support plate with a bobbin voltage greater than the lower voltage repair limit but less than or equal to the upper repair voltage limit, may remain in service if a rotating pancake coil inspection does not detect degradation. Steam generator tubes, with indications of ODSCC degradation with bobbin voltage greater than the upper voltage repair limit (as calculated in accordance with the methodology in Generic Letter 95-05 as L                    supplemented) will be plugged or repaired.'
l i          d)        As part of an inspection program to help ensurMhat additional degradation modec l                    are not occurring, all indications with bobbin coil voltage greater than 1.0 volt       ;
l                    will be inspected by rotating pancake coil (RPC) ?or the purpose of this l                    guidance, RPC inspection includes the use of comprable or improved inspection techniques.
e)        Section 2.0 of the attached Westinghnuse Report deceribes the applicability of the voltage-baced repair criteria and identifies those intersections that are excluded      l from the application of voltage-baced repair criteria. The criteria will be limited     j
,                    to mill annealed 600 allov tubes. excluding the Unit 2 thermally-treated tubes          l l                    from the application of the criteria. The Unit I steam generators have mill             l l                    annealed alloy 600 tubes only.
: 3. For each upcoming cycle, the projected end-of-cycle voltage distribution will be l
established based upon the previous end-of-cycle eddy current data. Based upon this distribution, postulated steam generator tube leakage during a steam line break will be estimated based on the guidance of Generic Letter 95-05. Projected leakage must remain H          below a level which results in offsite dose estimates remaining within the limits of              l l          10CFR100 and control room doses within GDC 19 limits. Should this estimation exceed j          the applicable dose limits, the highest voltage indications will be successively plugged          !
until the leakage estimation drops below the applicable dose limits. Projected steam generator tube leakage during a steam line break will be calculated as prescribed in Generic Letter 95-05.
L L   4. An overall tube burst probability during a postulated steam line break event will be calculated and compared to the threshold of 1 x 10 2 defined in Generic Letter 95-05.
: 5. Indications left in service as a result of application of the tube support plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages.
I-EAWIWIANRC-WK\TsC-98\0097. doc                                                             sTI: 30547247

NOC-AE-000097 Attochment 1 Page 5 of 5
: 6.       South Texas Proiect han determined no Unit 2 intersections would collanse or deform following a nostulated loss of coolant accident (LOCA) concurrent with a safe shutdoen L'            earthonalee (SSE) and therefore no Unit 2 tubes are excluded from the application of the
[             voltage-baced repair criteria for these reasons. Section 2.0 of the attached Westinghnuse l             Report describes those Unit 2 intersections that will be excluded from the annlication of the voltage haced repair criteria Ne in:: =:dc= fi!! M :=!rtd f:c . :pp!l= den cfi:
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              =!!:p= = d:fc :. fc!!cnin;; p=t" 'ed !c= cf =c!=: =:id::: (L^CA) :c=0.x::
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: 7.       If an unscheduled mid-cycle inspection is performed, the mid-cycle repair limits apply l             instead of the limits identified in Technical Specifications 4.4.5.4.a.12.a through L             4.4.5.4.a.12.c. The mid-cycle repair limits will be determined from the equations for          j mid-cycle repair limits in Generic Letter 95-05, Attachment 2, page 3 of 7.                    j L
Implementation of these mid-cycle repair limits.should follow the same approach as in          '
Technical Specifications 4.4.5.4.a.12.a through 4.4.5.4.a.12.c.
: 8.       All intersections with interfering signals greater than 1.0 volt from copper deposits with dents exceeding 5.0 bobbin volts, or with mixed residual signals, will be inspected with an RPC probe. Any indications found will be plugged or repaired.
i
: 9.       The current Technical Snecification renortmo reonirements of Section 4.4.5.5.d.
_ including several of those recommended by Generic Letter 95-065. are anolicable to Umt 2.
: 10.      Two non-technical changes are necessary to provide for a grammatical correction and to delete tube diameter information not applicable to South Texas that was introduced by the Generic Letter 95-05 Samnle Technical Soecifications.
I 1. Generic Letter 95-05 states that the probability of detection of ODSCC flaws (POD) should be 0.6 or an alternative approved by the NRC. This pronosed amendment NRC                l approval of an alternative POD as discussed in the attached Weginnhouse renortIsee nace 10).
i STP inspections are consistent with Generic Letter 95-05, Section 3, " Inspection Criteria."             i l
l E:\WP\NL\NRC-WK\TsC.98\0097. doc                                                          sTI: 30547247 l i
I

ATTACHMENT 2 SAFETY EVALUATION E:\%7NL\NRC-WK\TSC-9810097. DOC              STI: 30547247

NOC-AE-0097 Attachment 2 Page1of5 j
SAFETY EVALUATION i
In the development of the voltage-based repair criteria, draft Reg Guide 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes," and draft Reg Guide 1.83, Rev.1, " Inservice Inspection of PWR Steam Generator Tubes," are used as the bases for determining that steam generator integrity considerations are maintained within acceptable limits. Draft Reg Guide 1.121 describes a method acceptable to the NRC staff for meeting General Design Criteria (GDC) 14,15,31, and 32. The               J probability and consequences of steam generator tube rupture are reduced by determining the limiting safe conditions of degradation of steam generator tubing, beyond which tubes with unacceptable degradation, as established by inservice inspection, would be removed from service by plugging _or reair. This regulatory guide uses safety factors on loads for tube burst that are consistent with the requirements of Section III of the ASME Code. For the degradation occurring in the steam generator tube support plate elevations, tube burst criteria are inherently satisfied during normal operating conditions by the presence of the tube support plate. The presence of the tube support plate enhances the integrity of the degraded tubes in that region by precluding tube deformation beyond the diameter of the drilled hole. It is not certain whether the tube support plate would function to provide a similar constraining effect during accident condition loadings. Therefore, no credit is taken in the development of the voltage-based repair criteria for the presence of the tube support plate during         ;
accident condition loadings. Conservatively, based on the existing database, burst testing shows that      i the safety requirements for tube burst margins during both normal and accident condition loadings
                                                                                                           )
can be satisfied at end-of-cycle conditions with bobbin coil signal amplitudes of approximately 5.45       l volts for tube support plates and 4.47 volts for the flow distribution baffle, regardless of the depth of  ;
tube wall penetration degradation. Generic Letter 95-05 requirements for use of the latest NRC-            l approved database will be met and will ensure plant safety for future cycle structural repair limits.      !
Reg Guide 1.83, Rev.1, describes a method acceptable for implementing GDC 14,15,31, and 32                 '
through periodic inservice inspection for detection of significant tube wall degradation. STP Unit 2 is   ;
applying for voltage-based repair criteria for flow distribution baffle intersections.
For the voltage-based repair criteria developed for the steam generator tubes, no leakage is expected      i during normal operating conditions even with the presence of through-wall degradation. This is because the stress corrosion cracking occurring in the tubes at the support plate elevations in the steam generators are short, tight, axially oriented microcracks separated by ligaments of material.
Relative to the expected leakage during accident condition loadings, the limiting event with respect to   ,
primary-to-secondary leakage is a postulated steam line break event.
E:\%hNL\NRC WK\TsC-98\0097. Doc                                                           STI: 30547247

I NOC-AE-0097 Attachment 2 Page 2 of 5 The following items support this proposed license amendment.
: 1.      . Chemistry STP has undertaken steps to help mitigate steam generator tubing corrosion. Plant design was upgraded during construction to:
e       add a full flow feedwater deaerator for dissolved oxygen control,      ,
             .       add cation condensate polishers in addition to the full flow mixed bedbondensate polishers, e       double the capacity of the steam generator blowdown system to 1% of main steam flow, e       remove copper components from the secondary system, and                                  ,
e       use all volatile treatment                                                               I During the past three years, attemate amine pH control was implemented to reduce iron transport. Current information included in the EPRI Secondary Chemistry Guidelines is used to monitor the effectiveness of the chemistry program.                                           '
: 2.         Steam Generator Leakaoe Monitoriny i
Steam generator leakage monitoring employs a sampling program in conjunction with
         . radiation monitors permanently installed on the condenser air removal system (RT8027), the unit vent monitor (RT8010), the steam generator blowdown flash tank (RT8043), and employing N-16 primary-to-secondary leak monitors permanently installed on each of the four main steam lines (RT8130B, RT8131B, RT8132B, and RT8133B). The STP program -
for detection and mitigation of steam generator tube leak events was upgraded earlier in response to industry lessons learned such as Information Notice 91-43. The program for early
          ' leak detection provides for prompt detection and response, thus minimizing the likelihood of a steam generator tube rupture event. In addition to the monitors described below, supplementary monitors which are less sensitive to small leaks have been provided on each of _     l the four main steam lines and on the four steam generator blowdown lines. These are                l provided primarily for detection of a steam generator tube rupture event and are not discussed l
in detail.
                                                                                                            / 1 Samnling                                                                                        j Each steam generator is sampled for various purposes, including the detection of tube leaks and determination of secondary specific radioactivity.
E:\Wl9ft\NRC-WK\TsC-98\0097. doc                                                              sTI: 30547247 i

NOC-AE-0097 Attachment 2 Page 3 of 5 '
Steam Generninr Blowdown (SGBD) Radiation Moniinr The SGBD radiation monitor continuously checks the steam generator blowdown flash tank effluent to provide indication and alarms locally and in the control room. The radiation
             . monitor detects water activation products as well as corrosion activation products and fission products. It is sensitive to leakage as low as five gallons per day. An alert or high alarm would be an indication of a primary-to-secondary leak.
CnnAancar Air Remnval System Radiation Moniinr The condenser air removal system is provided with a radiation monitor which continuously monitors the effluent line from the condenser vacuum pump. This monitor is designed to detect low levels ofnoble gas radioactivity and is sensitive to leaks as low as five gallons per_ -
day. An alarm from this detector indicates a primary-to-secondary system leak.
Unit Vent Monitor The unit vent is provided with a radiation monitor which samples the plant vent stack prior to discharge to the environment and monitors for particulates, iodine, and noble gases. The unit vent monitor provides for sampling ofplant effluents in compliance with NUREG-0737, Item II.F.1.
N-16 Radiation Monitors The N-16 gamma detectors provide continuous indication ofindividual steam generator                 ;
primary-to-secondary leakage. The detectors provide real time indication in the control room        '
of steam generator leak rate in gallons per day when reactor power is 225%. The N-16 monitors are reactor power compensated for accurate leakage trending during power level             1 reductions and increases. A recorder monitors the N-16 detector readings and provides a             I trend recording of steam generator leak rate. The N-16 monitors alarm in the cold chemistry         :
lab, from which they are controlled, while monitor readings are continuously available in the control room via the plant computer.
Station Response to a Steam Generator Tube I eak Abnormal radiation in a steam generator indicates primary-to-secondary leakage. This can be shown by trends or alarms on main steam line N-16 monitors, the condenser vacuum pump effluent monitor, the steam generator blowdown radiation monitor, or from chemistry samples. A large leak would be indicated by feedwater flow being less than steam flow,              i decreasing feed flow, a mismatch in charging and letdown flow, or decreasing feed regulating        !
valve position in conjunction with a stable steam generator level. These symptoms, however,         j would 'nore likely be noticed with a tube rupture event. Procedures provide actions to              {
mitigate the entire spectrum of steam generator tube leaks from the threshold of detectability up to a steam generator tube rupture event.
  ' E:\WP\NL\NRC-WKYTSC.98\0097. doc                                                           STI: 30547247 l

NOC-AE-0097 Attachment 2 Page 4 of 5 Upon any confirmed indication ofleakage, the frequency of monitoring and sampling is
             ~  increased in a manner proportionate to the severity of the leak. Additional confirmatory or diagnostic samples would be taken from the steam genvator blowdown and from the condenser air removal system effluent. Operators begin to closely monitor the N-16 monitors in the control room.
I Trainino -
The operator training program Ins been upgraded previously to reflect training scenarios based on actual plant response to previous industry steam generator leak events. Plant operators and chemical analysis technicians have been trained in the use of the N-16 monitors and in the upgraded station procedures for response to steam generator leaks.
Steam Generntnr Leale Detection Pronram Adeanaev The plant leak rate monitors and procedures provide the required indications and alarms to ensure reactor coolant system leakage is detected early, while the leakage rate is low. In addition, leakage verification is provided by chemistry procedures which provide alternate means of calculating and confirming reactor coolant system leakage. These procedures maximize assurance that leak evaluation and mitigation can occur before small leaks propagate to steam generator tube mpture events, i
: 3.       Eddy Current Test and the Data Analysis Guidelines The data acquisition and the analysis guidelines will be in accordance with Westinghouse Steam Generator Repon SG-98-01-004 for type of calibration, recording, and analysis requirements.
4,      FAdv Current Data Annivst Trainino and Ounlifications
            - All analysts will be qualified per SNT-TC-1 A and a minimum of 90% will be qualified as
               " Qualified Data Analyst." Supervisors are not. required to be " Qualified Data Analysts." All analysts must pass the site-specific data analysis course prior to beginning work. The site-specific data analysis course will sensitize the analysts to identify indications attributable to primary water stress corrosion cracking (PWSCC) and to recognize the potentH for PWSCC to occur at dented tube support plate intersections. This information will be co..utined in the curren " Steam Generator Eddy Current Data Analysis Training Manual" or its equivalent.
: 5.         Tube Pulls Removal of a minimum of four tube support intersections containing larger distorted support indications (DSIs)is planned for refueling outag: 2RE06. The results of Unit I destructive examination of eighteen tube suppon plate intersections from four tube pulls performed in         -
1993 and three tubes pulls performed in 1995 are included in Section 8 of BAW-10204-P, Rev. 3 (refer to letter ST-HL-AE-5395 dated 6/6/96). From these pull samples, one ~70%
through wall axial defect and one ~54% axial defect were burst tested. No other degradation E:\WP\NL\NRC-W\Tsc.98\0097. Doc                                                              sTI: 30547247

c NOC-AE-0097 Attachment 2 Page 5 of 5 morphologies, such as PWSCC, circumferential degradation, or significant IGA, were found
            ;in the support plate intersections. Helium leak testing indicated no through-wall penetrations and thus hot leak rate testing was not performedi This information has been incorporated as appropriate into the industry-wide tube pull database.
STP Unit 2 is committed to tube pulls as required by Generic Letter 95-05 or to participation in an industry tube pull program should such a program be approved by the NRC as allowed
           ' by Generic Letter 95-05.
The proposed amendment may preclude occupational radiation exposure that would
"                                                                                                           j otherwise be incurred by plant workers involved in tube plugging operations. Operation in       !
accordance with the proposed amendment would reduce loss of margin in reactor coolant flow through the steam generator and is therefore safety enhancing. Its implementation would assist in assuring that minimum flow rates are maintained in excess of that required for
  ,          operation at full power. Reduction in the amount of tube plugging can reduce the length of plant outages and reduce the time the steam generator is open to the containment environment during an outage. STP has determined that this methodology is applicable to        1 our steam generators and provides a safe and effective alternative to plugging.
: 6. Two non-technical changen are nececcarv to provide for a grammatical correction and to delete tube diameter information not anplicable to South Texas that was introduced by the Generic Letter 95-05 Samnle Technient Snecincations. These changes do not change the intent or the basis of the Technical Specifications.
The reporting reouirements for the voltage haced repair criteria currentiv reonired for Unit I are applied to Unit 2. This will ensure consistency between South Texas Units and promnt notification of the NRC prior to returning the steam generators to service.
: 7. The voltane-based repair criteria will be limited to mill annealed alloy 600 tubes. excluding the Unit 2 thermally-treated tubes from the application of the criteria. Although thermally treated allov 600 tube material is less succentible to stress corrosion crackino than mill
                                                                                          ~
annealed allov 600 tube material. the nostulated corrosion mornhology can not be adannately determined without a tube pull should ODSCC develop in these tubes. The Unit I steam generators have mill annealed allov 600 tubes only.
EM ML\NRC-WK\TsC-98\0097. doc                                                           sTI: 30547247

ATTACHMENT 3 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION E:\WP\NL\NRC-WK\TSC-98\0097. DOC              STI: 30547247

1 NOC-AE-000097 Attachment 3        j Page1of5 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION The following evaluation determines that the proposed amendment to the Technical Specifications involves no significant hazards consideration as defined in 10 CFR 50.92. The Unit I changes that involve wording changes to ensure consistency between both Units are administrative and have no effects that could imnact the determination below.
                                                                                                           )
: 1.        Operation of the facility in accordance with the proposed amendment would not                I involve a significant increase in the probability or consequences of an accident previously evaluated.                                                                        {'
Structural Considerations Industry testing of model boiler and operating plant tube specimens for free span tubing at room temperature conditions shows typical burst pressures in excess of 5000 psi for indications of ODSCC     !
with voltage measurements at or below the current structural limit of 5.45 volts. One model boiler specimen with a voltage amplitude of 19 volts also exhibited a burst pressure greater than 5000 psi.
Buret testing performed on one intersection pulled from STP Unit 1 in 1993 with a 0.51 volt             !
indication yielded a measured burst pressure of 8900 psi at room temperature. Burst testing perfonned on another intersection pulled from STP Unit 1 in 1995 with a 0.48 volt indication yielded a measured burst pressure of 9950 psi at room temperature.
The next projected end-of-cycle (EOC) voltage compares favorably with the current structural limit considering the voltage growth rate for indications at STP. Using the methodology of Generic Letter 95-05, the structural limit is reduced by allowances for uncertainty and growth to develop a beginning-of cycle (BOC) repair limit which should preclude EOC indications from growing in excess of the structural limit. The non-destructive examination (NDE) uncertainty to be applied per Generic Letter 95-05 is approximately 20%. The growth allowance will be 30%/EFPY or a STP Unit 2-specific growth rate, to be calculated in accordance with Generic Letter 95-05, whichever is greater.
Where the generator-specific growth rate exceeds both the Unit 2-specific average growth rate and 19%/EFPY, that generator-specific growth rate will be used for that generator. Each succeeding cycle upper voltage repair limit will also be conservatively established based on Generic Letter 95-05 methodology. By adcling NDE uncertainty allowances and a growth allowance to the repair limit, the structurallimit can be validated.
The upper voltage repair limit could be applied to bobbin coil voltages between the lower and upper repair limits to leave such indications in service independent of RPC confirmation. However, RPC-confirmed indications will be conservatively removed from service consistent with Generic Letter 95-05.
E;\WP\NL\NRC-WK\TsC-98\0097. DOC                                                       sTI: 30547247

I NOC-AE-000097 Attachment 3 Page 2 of 5 Leat nge Considerations As part of the implementation of voltage-based repair criteria, the distribution of EOC degradation indications at the TSP intersections has been used to calculate the primary-to-secondary leakage which is bounded by the maximum leakage required to remain within the applicable dose limits of 10CFR100 and GDC 19. This limit was calculated using the Technical Specification Reactor                    !
Coolant System (RCS)RGS Iodine-131 transient spiking values consistent with NUREG-0800.                     I Application of the voltage-based repair criteria requires the projection of postulated Main Steam Line      l Break (MSLBMSbB leakage based on the projected EOC voltage distribution from the beginning of cycle voltage distribution. Projected EOC voltage distribution is developed using the most recent EOC eddy current results and a voltage measurement uncertainty. Draft NUREG-1477 and Generic Letter 95-05 require that all indications to which voltage-based repair criteria are applied must be        l included in the leakage projection.
The projected MSLB leakage rate calculation methodology prescribed in Generic Letter 95-05 will be ured to calculate the EOC lealage. A Monte Carlo approach will be used to determine the EOC leakage, accounting for all of the bobbin coil eddy current test uncertainties, voltage growth, and an      i assumed probability of detection of 0.6. The fitted log-logistic probability ofleakage correlation will     l be used to establish the MSLB leak rate for each cycle. This leak rate will be used for comparison with a bounding allowable leak rate in the faulte ; loop which would result in radiological                  l consequences which are within the dose limits of 10CFR100 for offsite doses and GDC 19 for control room doses. Due to the relatively low voltage levels ofindications at STP to date and low voltage
. growth rates, it is expected that the actual calculated leakage values will be far less than this limit for ;
each successive cycle.                                                                                       l Other Considerations Those changes associated with grammatical corrections. deleting tube diameter information not app'icable to South Texas. and annlying the additional renorting reanirements to Unit 2. are administrative and do not involve a change to. or the operation of. any safety-related system.
Therefore, implementation of voltage-based repair criteria does not adversely affect steam generator
. tube integrity and the radiological consequences will remain below the limits of 10CFR100 and GDC 19. Operation of the facility in accordance with the proposed amendment would not result in any increase in the probability or consequences of an accident previously evaluated.
: 2.        Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.
Implementation of the proposed steam generator tube voltage-based repair criteria for ODSCC at the TSP intersections does not introduce any significant changes to the plant design basis. Use of the criteria does not provide a mechanism which could result in an accident outside of the region of the TSP elevations because the criteria do not apply outside the thickness of the TSPs. It is therefore E:\WRNUNRC-WKiTsC-98\0097. doc                                                            sTI: 30547247

?.
l NOC-AE-000097 Attachment 3 Page 3 of 5 expected that for all plant conditions, neither a single nor multiple tube rupture event would likely
: j. . occur in a steam generator where voltage-based repair criteria has been applied.
Specifically, STP UnM-has implemented a maximum leakage rate of 150 gpd per steam generator l
to help preclude the potential for excessive leakage during all plant conditions. The draR Reg Guide 1.121 criterion for establishing operational leakage rate limits governing plant shutdown is based upon leak-before-break (LBB) considerations to detect a free span crack before potential tube rupture as a result 0: fac!ted plant conditions. The 150 gpd limit is intended to provide for leakage detection and plant shutdown in the event of unexpected crack propagation outside the tube support plate thickness resulting in excessive leakage. DraR Reg Guide 1.121 acceptance criteria for establishing operating leakage limits are based on LBB considerations such that plant shutdown is initiated if permissible degradation is exceeded.
f Thus, the 150 gpd limit provides for plant shutdown prior to reaching critical degradation lengths.
Additionally, the leak-before-break evaluation assumes that the entire crevice area is uncovered during the secondary side blowdown of a MSLB. Typically, it is expected for the vast majority of intersections, that only partial uncovery will occur. Therefore, the proximity of the TSP will enhance the burst capacity of the tube.
Steam ;;nerator tube integrity is continually maintained through inservice inspection and primary-to-secandary leakage monitoring. Any tubes falling outside the voltage-based repair criteria limits l-   are removed from service.
Those changes acenciated with grammatical corrections. deletino tube diameter information not applicable to South Texae and applying the additional renortino reauirements to Unit 2. are administrative and do not involve a chance to. or the operation of. any safetv-related system.
Therefore, operating the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.
: 3.         Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.
The use of the voltage-based bobbin probe for dispositioning ODSCC degraded tubes within TSP intersections is demonstrated to maintain steam generator tube integrity in accordance with the l    requirements of draR Reg Guide 1.121. Dran Reg Guide 1.121 describes a method acceptable to the NRC staff for meeting GDCs 14,15,31, and 32 by reducing the probability or the consequences of z steam ' generator tube rupture. This is accomplished by determining the limiting conditions of i
degradation of steam generator tubing, as established by inservice inspection, for which tubes with unacceptable degradation are removed from service. Upon implementation of the criteria, even under the worst case conditions, the occurrence of ODSCC at the TSP elevation is not expected to lead to a steam generator tube rupture event during normal or faulted plant conditions. The EOC
[
distribution ofindications at the TSP elevations for each successive cycle will be confirmed to result in acceptable primary-to-secondary leakage during all plant conditions.
1 EAWPNDNRC-WK\TsC-98\0097. DOC                                                              sTI: 30547247 l
l l

NOC-AE-000097 Anachment 3         1 Page 4 0f 5 i
In addressing the combined effects ofloss of coolant accident (LOCA) and safe shutdown                    {
earthquake (SSE) on the steam generators, as required by GDC 2, it has been determined that tube collapse may occur in the steam generators at some plants. This is not the case at STP Unit 2 as the
                                                                                                            )
: TSPs do not become sufficiently deformed as a vsult oflateral loads at the wedge supports at the l
periphery of the plate due to the combined efTects i f the leak-before-break-limited LOCA rarefaction   .j wave and SSE loadings to affect tube integrity.
l Because the leak-before-break methodology is applicable to the STP reactor coolant loop piping, the       ;
probability of breaks in the primary loop piping is sufficiently low that they need not be considered      !
in the structural design of the plant. Implementation practices using the bobbin probe voltage based tube plugging criteria bounds Reg Guide 1.83, Rev.1, considerations by:                                 1
: 1)     Using enhanced eddy current inspection guidelines consistent with those used by EPRI in developing the correlations. This provides consistency in voltage                 l normalization.                                                                            4
: 2)      Performing a 100% bobbin coil inspection for all hot leg tube support plate intersections and all cold leg intersections down to the lowest cold leg tube support     !
plate with known ODSCC indications at each cycle. The determination of the tube support plate intersections having ODSCC indications shall be based on the performance of at least a 20% random sampling of tubes inspected over their full          l length, and                                                                               ;
i 3)-     Incorporating rotating pancake coil inspection for all tubes with bobbin voltages greater than 1.0 volt. This further establishes the principal degradation morphology as ODSCC.
Implementation of voltage-based repair criteria at TSP intersections will decrease the number of           1 tubes which must be repaired at each subsequent inspection. Since the installation of tube plugs to remove ODSCC degraded tubes from service reduces the RCS flow me;%, voltage-based repair criteria implementation will help preserve the margin of flow.
For each cycle the projected EOC primary-to secondary leak rate allowed is bounded by a leak rate which limits the radiological consequences of a EOC MSLB to within the dose limits of 10CFR100 for offsite doses and 10CFR50 Apnendix A General Design Criteria (GDCEDC 19 for control l
room doses. Therefore, this change does not involve a significant reduction in the margin to safety.
The assessment of radiological consequences of an assumed steam line heak applicable to STP Unit I was provided in Attachment 2 to ST-HL-AE-5359 on May 2,1996. The submittal was made in response to questions from the Emergency Preparedness and Radiation Protection Branch and is applicable to Unit 2 as well. The staficoncluded that the thyroid doses for the Exclusion Area Boundarv (EABEAR, Low Popnhtion Zone (LPZ%p;r,, and control room are within the acceptance criteria.
E:\WP\NL\NRC WK\TsC-98\0097. DOC                                                          sTI: 30547247

NOC-AE-000097 Attachment 3 Page 5 of 5 Those changes associated with grammatical corrections. deleting tube diameter information not applicable to South Texac and applying the additional renorting requirements to Unit 2. are administrative and do not involve a change to. or the oneration of. any safetv-related system.
It is therefore concluded that the proposed license amendment request does not result in a significant reduction in the margin of safety as defined in the plant Final Safety Analysis Report or Technical Specificaticins.
)
E\WPNL\NRC-WK\TSC-98\0097. DOC                                                           sTI: 30547247}}