View source for ST-HL-AE-5689, Application for Amend to License NPF-76,incorporating Revs to TS 3.7.1.6, Atmospheric Steam Relief Valves, to Ensure Automatic Feature of SG Power Operated Relief Valve Remains Operable During Modes 1 & 2

{{Adams
| number = ML20217M520
| issue date = 08/18/1997
| title = Application for Amend to License NPF-76,incorporating Revs to TS 3.7.1.6, Atmospheric Steam Relief Valves, to Ensure Automatic Feature of SG Power Operated Relief Valve Remains Operable During Modes 1 & 2
| 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 = ST-HL-AE-5689, NUDOCS 9708250025
| package number = ML20217M525
| document type = OPERATING LICENSES-APPLIATION TO AMEND-RENEW EXISTING, TEXT-LICENSE APPLICATIONS &amp; PERMITS
| page count = 10
| project = 
| stage = Request
}}

=Text=
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The Light cIl6uston o mp      a ny S uth Texas Project Electric Generating l$ighting & Power Station P. O. Box 289 Wadsworth, Texas 77483 August 18,1997 ST-llL-AE-5689 File No.: G20.02.01 10CFR50.90 10CFR50.92 10CFR51 STI: 30332447 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 South Texas Project Units 1 & 2 Docket Nos. STN 50-498, STN 50-499 Proposed Amendment of Technical Speci0 cation 3,7.1.6. Atmospheric Steam Relief Valves

==Reference:==
Letter from D. A. Leazar to NRC Document Control Desk,"10CFR50.46 30 Day Report of Signincant Changes to the Accepted Emergency Core i                           Cooling System Model", dated July 17,1997,(ST-IIL-AE 5698)
I The South Texas Project proposes to amend Facility Operating Licenses NPF-76, Unit I and NPF-80, Unit 2 by incorporating revisions to Technical Specification 3.7.1.6, Atmospheric Steam Relief Valves, to ensure the automatic feature of the Steam Generator Power Operated Relief Valve remains operable during Modes 1 and 2. In addition, the proposed change adds a surveillance that requires a Channel Calibration on the Steam Generator Power Operated Relief Valve be performed every 18 months.
A reanalysis of the Small Break Loss of Coolant Accident (SBLOCA) event at the South Texas Project determined that a previous assumption of no safety injection flow into the broken loop was not conservative. With safety injection flow into the broken loop considered, analysis results indicated that the Peak Clad Temperature acceptance                       miliiii.
limit of 10CFR50.46 could be exceeded unless cacdit is taken for the automatic actuation                    @
of the Steam Generator Power Operated Relief Val'ces. If the automatic feature of the Steam Generator Power Operated Relief Valve remains operable during Modes 1 and 2, g
gir the analysis results show a calculated Peak Clad Temperature well below the acceptance                      E' limit of 10CFR50.46 and comparable to the results currently described in the Updated Final Safety Analysis Report. The change in the reanalysis of the SBLOCA event and the E
g.
action taken to show compliance with 10CFR50.46 requirements y as reimned to the                            -
Nuclear Regulatory Commission by the referenced letter on July 17,1997. Tre Steam Generator Power Operated Relief Valves are Class IE powered and have safety grade                       ,
                                                                                                                  / 00)         )
Project Manager on BehaH of the Participants in the South Texas Project   sTI: 30332447 9708250025 970818 PDR      ADOCK 05000498
       $                        PDR o

4; Houston Lighting & Power Company South Texas Project Electric Generating Station ST-HL-AE-5689 File No.: G20.02.01 Page 2
                - automatic actuation. Pending approval of this proposed Technical Specification Amendment, administrative controls have been established to require the automatic actuation of the Steam Generator Power Operated Relief Valves when Units 1 and 2 are in Modes 1 and 2.
l                        The South Texas Project has reviewed the attached proposed amendment pursuant to 10CFR50.92 and determined that it does not involve a significant hazards consideration in addition, the South Texas Project 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 South Texas Project Plant Op: rations Review Committee and Nuclear Safety Review Board have reviewed and approved the proposed change.
The required affidavit, along with a Safety Evaluation and No Significant Hazards Consideration Determination associated with the proposed changes, and the marked-up Technical Specification pages are included as attachments to this letter.
In accordance with 10CFR50.91(b), South Texas Project is providing the State of Texas with a copy _of this proposed amendment.
if you should have any questions concerning this matter, please call Mr. A. W Harrison at (512) 972-7298 or me at (512) 972-8787, v
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T. H. loni Vic re       nt, N lear Engin        ing KJT/
                ' Attachments: 1.          Affidavit -
: 2.       Safety Evaluation and No Significant Hazards Consideration Determination -
: 3. -     Proposed Change to Technical Specification 3.7.1.6 STI: 3t 132447

Houston Li                                                        ST IIL AE 5689 South TexProject
:ghting  & Power Compen Electric Generat ng Station                   File No.:  G20.02.01
                   -                                                      Page 3 Ellis W hierschoff                              Rufus S. Scott Regional Administrator, Region IV               Associate General Counsel U. S. Nuclear Regulatory Commission             llouston Lighting & Power Company 611 Ryan Plaza Drive, Suite 400                 P. O. Box 61067 Ar'Qton, TX 76011 8064                          Ilouston, TX 77208 Thomas W. Alexion                               Institute of Nuclear Power l
Project hianager, hiall Code 13113              Operations Records Center U. S. Nuclear Regulatory Commission             700 Galleria Parkway Washington, DC 20555 0001                       Atlanta, GA 30339 5957 David P. Loveless                               Dr. llertram Wolfe Sr. Resident inspector                           15453 Via Vaquero
;       c/o U. S. Nuclear Regulatory Comm.              hionte Sereno, CA 95030 l       P. O. Ilox 910 Bay City TX 77404-0910                          Richard A. Ratliff Bureau of Radiation Control J. R. Newman, Esquire                          Texas Department of llealth h1 organ, Lewis & llockius                      1100 West 49th Street 1800 h1 Street, N.W.                           Austin, TX 78756 3189 Washington, DC 20036 5869 J. R. Egan, Esquire hi. T. liardt/W. C. Gunst                      Egan & Associates, P.C.
City Public Service                            2300 N Street, N.W.
P. O. Box 1771                                 Washington, D.C. 20037 San Antonio, TX 78296 J. C. Lanier/hi. B. Lee                         U. S. Nuclear Regulatory Commission City of Austin                                  Attention: Document Control Desk Electric Utility Department                    Washington, D.C. 20555-0001 721 Barton Springs Road Austin, TX 78704 Central Power and Light Company ATTN: G. E. Vaughn/C. A. Johnson P. O. Box 289, Mail Code: N5012 Wadsworth, TX 77483

e S T ill.AE 5689 1
1 1
ATTACllMENT I AFFIDAVIT STI: 30332447

i Attachment 1
_ ST llL-AE 5689 UNITED STATES OF AMERICA                                                               i NUCLEAR REGULATORY COMMISSION in the Matter                                        )
                                                                   )
South Texas Project, et al.,                         )                          Docket Nos 50 498
                                                                   )                                      50-499 South Texas Project Units 1 & 2                      )
AEEIDAVIT l
1 T.11. Cloninger, being duly sworn, hereby depose and say that I am Vice President, Nuclear Engineering, of South Texas Project; that I am duly authorized to sign                              ,
and file with the Nuclear Regulatory Commission the attached proposed amendment to Technical Specification 3.7.1.6; that I am familiar with the content thereof; and that th matters set forth therein are true and correct to the best of my knowledge and be                    .                 4 g                     .
T.11. Cl in Vice esi nt, Nu       arl' gineering STATE OFTEXAS COUNTY OF MATAGORDA                                   )
Subscribed and sworn to before me, a Notary Public in and for the State of Texas, this /f* day of Agasf",1997.
g-  [,W e           _.
ary Public hi and for the
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State of Texas
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Attachment 2 ST llL-AE 5689 Page1of5 SAFETY EVALUATION AND NO SIGNIFICANT llAZARDS CONSIDERATION DETERMINATION l                                                                                    llackground i                                         During long term planning and preparation for steam generator replacement at the South Texas Project, an analysis was completed to assess the impact of safety injection now into the broken loop for a Small Break Loss of Coolant Accident (SDLOCA) event. The analysis results identified the potential for the calculated Peak Clad Temperature to exceed the 2200 F limit of 10CFR50.46 for Units I and 2. An additional analysis demonstrated that by taking credit for the automatic actuation of the Class IE powered Steam Generator Power Operated Relief Valves, the Peak Clad Temperature would be reduced to considerably below 2200*F. As a remedial action, administrative controls have been established to require the automatic aciuntica of the Steam Generator Power Operated Relief Valves when Units 1 and 2 are in Modes 1 and 2. Since the automatic actuation of the Steam Generator Power Operated Relief Valves is now found necessary to mitigate the consequences of a SBLOCA event r.nd meet the requirements of 10CFR50.46, the South Texas Project submits this Technical Specification Amendment to add this Limiting Cr,adition for Operation in accordance with criterion (3) of 10CFR50.36(c)(2)(ii).
When the South Texas Project was or;ginally licensed, the SDLOCA event analysis assumed no safety injection How into the broken loop. Subsequent analysis of the SDLOCA event showed this assumption was not conservative. Safety injection flow into the broken loop should be considered. The relatively cool safety injection Dow subcools the fluid in the cold leg where the break is assumed to occur. The density of the Huid in the Reactor Coolant System cold leg increases which causes the mass now rate out of the break to increase. As a result, the water mass in the Reactor Coolant System decreases so that the core becomes uncovered for a longer period of time than the condition when no now into t'ne broken loop is assumed. Consequently, an increase in Peak Clad Temperature occurs as compared to the result in the original analysis.
Analysis performed on generic 4-loop plants showed a 150 F penalty to account for the subcooling ef fect on the cold leg due to safety injection flow into the broken loop. This penalty, when added with other penalties und to the calculation of record, resulted in a calculated Peak Clad Temperture to 2158*F. For a typical 4-loop plant, approximately one-fourth of the safety injection flow enters the broken loop because the safety injection sTI: 30332447

Attachment 2 ST llL-AE-5689 Page 2 of 5 system is cross headered to each loop. In the South Texas Project design, the redundant safety injection trains are not cross headered; each train only flows into its respective loop.
Therefore, after assuming single failure on one of the three safety injection trains, one half of the available safety injection flow enters the broken loop and considerably more subcooling of the fluid in the broken loop occurs. Additional analysis using a South Texas           i Project specific model showed that the 150*F penalty was not sufficient. By considering the unique non header design of the South Texas Project's safety injection system, the Peak Clad Temperature could exceed the 10CFR50.46 limit of 2200 F.
During the SBLOCA event, the pressure in the Reactor Coolant System equilibrates to just above the saturation pressure of the Steam Generators until the reactor coolant loop I
I seal clears. The saturation pressure of the Steam Generators is currently determined by the lowest setpoint of the Main Steam Safety Valves. The current analysis assum:s the lowest Main Steam Safety Valve setpoint is 1285 psig with a 3% uncertainty of 39 psi for a setpoint of 1324 psig. At this pressure, safety injection now into the Reactor Coolant System is much lower than the break Dow which results in a net loss of Reactor Coolant System inventory and a more protracted core uncovery, if the saturation pressure of the Steam Generators is assumed to be determined by the 1225 psig setpoint with uncertainties associated with the automatic actuation of the Steam Generator Power Operated Relief Valve, Steam Generator and Reactor Coolant System pressure would be lower. Safety injection now into the Reactor Coolant System would increase which significantly lowers the calculated Peak Clad Temperature to a value below the limit of 10CFR50.46.
Proposed Change Description The proposed change revises the Limiting Condition for Operation of Technical Specification 3.7.1.6 to ensure the automatic feature of the Steam Generatur Power Operated Relief Valve remains operable during Modes 1 and 2. This will allow mitigation of the consequences of a SBLOCA Nent and meet the requirements of 10CFR50.46, in addition, the proposed change adds nrvei!!ance that requires a Channel Calibration to include verincation of automatic act. - .on at the 1225 psig setpoint on the Steam Generator Power Operated Relief Wlve be performed every 18 months. This surveillance interval is consistent with assumed uncertainties in the safety analysis to ensure the Steam Generator Power Operated Relief Valves will perform their intended function.
sTI: 30D2447

4 Attachment 2 ST HL-AE-5689 Page 3 of 5 Safety Evaluation                                           :
l       The NOTRUMP computer code is used in the analysis of SBLOCAs in the Reactor i
Coolant System. The modeling of Steam Generator secondary side atmospheric relief capability has always teen an impodant feature of the Westinghouse SBLOCA Evaluation Model. Atmospheric reliefis significant to the depressurization of the primary and secondary system during a SBLOCA. This depressurization directly influences the rates of break flow from the reactor vessel and safety injection into the reactor vessel. The modeling of safety grade Steam Generator Main Steam Safety valves is standard for all NOTRUMP analysis with Westinghouse Nuclear Steam System Supply designs.
A Peak Clad Temperature analysis for the SBLOCA event was performed using the currently approved NOTRUMP Evaluation Model, WCAP-11232, referenced in the South Texas Project Updated Final Safety Analysis Repon. The modelincluded asymmetric Emergency Core Cooling System and Auxiliary Feedwater System flows unique to the South Texas Project. The model also credits the Class lE-powered Steam Generator Power Operated Relief Valves for the South Texas Project. The modeling of the Steam Generator Power Operated Relief Valves was performed consistent with the modeling of secondary atmospheric relief valves as described in the NOTRUMP evaluation model, WCAP-10054 P-A. Since the analysis takes credit for the the automatic feature of the Steam Generator Power Operated Relief Valve, a single failure of a Steam Generator Power Operated Relief Valve has been considered in the single failure analysis for the SBLOCA event.
The plant specine reanalysis reconsidered various combinations of'oreak locations and limiting single failure scenarios and break locations that are unique to the South Texas Project. The results of the reanalysis show that the Peak Clad Temperature with t. I penalties considered is 1860 F which is well below the 10CFR50.46 acceptance limit of 2200 F.
S1l: 30332447

Attw4dnent 2 ST!!!L-AE-5689 Page 4 of 5 No Significant llazards Consideration Determination The South Texas Project proposes to revise Technical Specification 3.7.1.6 to ensure the automatic feature of the Steam Generator Power Operated Relief Valve remains operable during Modes 1 and 2. The South Texas Project has evaluated this proposed amendment l        and determined that it involves no significaat hazards considerations based on the following:
A.      The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The methodologies used in the accident analyses remain unchanged. The automatic actuation of the Steam Generator Power Operated Relief Valves is not a
,                new design feature. The effects of the inadvertent opening of a Steam Generator Power Operated Relief Valve are currently analyzed as described in Section 15.1.4 of the Updated Final Safety Analysis Report. The radiological consequences for the SDLOCA event presented in the Updated Final Safety Analysis Report remain unchanged. The calculated Peak Clad Temperature remains substantially below the 2200 F acceptance limit of 10CFR50.46.
B. The proposed change does not create the possibility of a new or dilTerent kind of accident from any accident previously evaluattd.
The automatic actuation of the Steam Generator Power Operated Relief Valves is not an accident initiator for the SBLOCA event. The automatic actuation of the Steam Generator Power Operated Relief Valves currently exists at the South Texas Project and is not a new design feature. The description of the Steam Generator Power Operated Relief Valves currently exists in the Updated Final Safety Analysis Report. This change does not represent a change to the facility and does not affect the safety functions and reliability of systems, structures, or components in any new manne - Operating procedures have a temporary administrative control to ensure the automatic actuation of the Steam Generator Power Operated Relief Valves remains operable in Modes I and 2. This condition will become permanent with the approval of this Technical Specification Amendment proposal.
STI: 30332447

Attachment 2 ST llL AE 5689 Page 5 of 5              ,
C. The proposed change does not involve a significant reduction in a margin of safety.
The proposed change results in the calculated Peak Clad Temperature remaining well below the acceptance limit of 10CFR50,46 and comparable to the results currently described in the Updated Final Safety Analysis Report.                     ,
Therefore, the South Texas Project has concluded that the proposed change does not involve significant hazards considerations.
Implementation Schedule Upon Nuclear Regulatory Commission approval of this proposed Technical Specification Amendment, the South Texas Project requests 30 days to implement the amendment.
STI: y332447}}