ML20210J995

Revised License Amend Request 269 for License DPR-50, Revising TMI-1 UFSAR Section 14.1.2.9 Environ Dose Consequences for TMI-1 Steam Line Break Analysis

ML20210J995
Person / Time
Site:	Three Mile Island
Issue date:	08/14/1997
From:	Langenbach J GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20210K000	List: ML20210J995 ML20210K001
References
6710-97-2345, NUDOCS 9708190104
Download: ML20210J995 (8)

Text

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4# 4 l GPU Nuclear, Inc.

( Route 441 south NUCLEAR 'jy,R" By*,*,%,,,

Tel717 944 7621 6710-97-2345 August 14, 1997 U.S. Nuclear Regulatory Commission l Attn: Document Control Desk Washington, DC 20555 Gentlemen:

Subject:

Three Mile Island Nuclear Station, Unit 1 (TMI-1)

Docket No. 50-289 Operating License No DPR-50 License Amendment Request No. 269 - Revised Steam Line Break Accident Analysis Dose Consequence in accordance with 10 CFR 50 4(b)(1), enclosed is License Amendment Request No. 269.

The purpose of this License Amendment Request is to revise the TMI-l Updated Final Safety Analysis Report (UFSAR) Section 14.1.2.9-Steam Line Break analysis to include the environmental dose consequenccs associated with postulated accident induced steam generator tube leakage not previously analyzed. The revised environmental dose consequences for the TM1-1 Steam Line Break analysis would be increased above the values previously reviewed by the NRC, but they continue to be below the limits containeJ in 10 CFR 100 for the exclusion area boundary and low population zone as specified in Standard Review Plan 15.1.5, Appendix A. It is our understanding that upon NRC review and approval of the proposed FSAR change, NRC authorization of the change is issued by the addition of a Lic2nse Condition and a new License l Appendix describing the authorized FSAR change. Accordingly, Enclosure 1 also includes a proposed License Condition and new Appendix pages, which are consistent with recent FSAR change request approval amendments issued to other licensees.

Using the standards in 10 CFR 50.92, GPU Nuclear has concluded that these proposed changes do not constitute a significant hazards consideration, as described in the enclosed analysis performed in accordance with 10 CFR 50.91(a)(1). Also enclosed is the Certificate of Service for this request certifying service to the chief executives of the township and county in which the I facility is located, as well as the designated official of the Commonwealth of Pennsylvania, Bureau of Radiation Protection. }q 9708190104 970814 "

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6710-97 2345 License Amendment Request No. 269 Page 2 of 2 l- Approval of this license amendment to authorize the identified FSAR changes is re-quested prior to September 29,1997, to support use of steam generator tube inspection criteria for kinetically expanded regions during the upcoming 12R outage.

Sincerely, khish '

James W. Langenbach Vice President and Director, TM1

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Enclosures:

(1) TMI-l License Amendment Request No. 269 Safety Evaluation, No Significant Hazards Consideration, License Revised Pages t

(2) Affected TMI l Updated Final Safety Analysis Report Pages (3) Certificate of Service for TM1-1 License Amendment Request No 269 cc: Administrator, Region 1

- TMI Senior Resident inspector TMI-l Senior Project Manager

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htETROPOLITAN EDISON COh1PANY IERSEY CENTRAL POWER & LIGliT COhiPANY AND PENNSYLVANIA ELECTRIC COhiPANY THREE hilLE ISLAND NUCLEAR STATION, UNIT 1 Operating License No. DPR-50 Docket No. 50-289 License Amendment Request No. 269 COhih10NWEALTH OF PENNSYLVANIA )

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COUNTY OF DAUPillN )

This License Amendment Request is submitted in support of Licensee's request to change the License Conditions and add License Appendix B to Operating License No. DPR-50 for Three hiile Island Nuclear Station, Unit 1. As a part of this request, proposed replacement pages for Operating License DPR-50 are also included. All statements contained in this submittal have been reviewed, and all such statements made and matters set forth therein are true and correct to the best of my knowledge.

GPU NUCLEAR INC.

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L' BY: _ pee President and Director (Thil Sworn an Subscrib to before me this/// ay of ,1997.

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• Suzanne C. M;JcA Nomy Pub'ic Londonderry Twp.. DauoNn County M/Cemrrutsion E,xpirss Nov. 22,1HD Member, Pennsy!vano A; soc @on of timet

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4 9 ENCLOSURE 1 TMI-1 License Amendment Request No. 269 Safety Evaluation No Significant Hazards Consideration and Proposed License Revised Pages l

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• Ericlosure 1 6710 97-2345 Page1of4

1. license Amendment Request No. 269 GPU Nuclear requests that the following changed replacement page be inserted into the existing License:

Revised License Page: Page 7 GPU Nuclear requests that the following License page be added:

j Add: Appendix B These pages are attached to this enclosure.

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11. Reason for Changg 1

The purpose of this License Amendment Request is to revise the TMI-l Updated Final Safety Analysis Report (UFSAR) Section 14.1.2.9 - Steam Line Break Analysis, to include the environmental dose consequences associated with postulated accident induced steam

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generator tube leakage. The accident induced steam generator tube leakage is conservatively postulated to occur through potential tube defects. TM1-1 plans to perform eddy current testing (ECT) of the once-through steam generator (OTSG) tubes in the kinetically expanded tube-to-tubesheet joint area in the upcoming 12R outage (September 1997). Potential tube wall defects identified during the ECT in this area may be dispositioned as acceptable based on the criteria previously submitted to the NRC (GPU Nuclear letter to NRC 6710-97-2348, dated August 8,1997). This inspection criteria ensures that the OTSG tubes remain structurally capable of sustaining normal operating and accident induced loads. However, the hypothetical steam line break accident induced loads have the potential to relieve the contact pressure within the kinetically expanded tube-to-tubesheet joints. Therefore, the existing TMI-l steam line break dose consequences are being revised to conservatively account for any postulated primary-to-secondary leakage.

111. Safety Evaluation Justifying Change The proposed change involves a change to the dose consequences presently contained in the TMI-l UFSAR, Section 14.1.2.9 - Steam Line Break Analysis, d'ie to a revised assumption ,

of accident induced primary-to-secondary leakage through potential tube flaws. I.eakage is conservatively postulated to occur through potential tube wall flaws as a result of hypothetical accident induced axial loads on the OTSG tubes. These loads could also relieve the contact pressure between the kinetically expanded tube-to-tubesheet joints, thus providing a potential primary-to-secondary leakage path post-accident. The steam line break accident induced tube loads will not result in a tube break or separation from the tubesheet. Therefore, the existing UFSAR Section 14.1.2.9 acceptance criteria that no steam generator tube break or separation from the tubesheet will occur is maintained. The OTSG tube stmetural analysis supporting the ;tmetural integrity of OTSG tubes with potential flaws in the kinetic expansion area is described in MPR Associates, Inc., Report

' Enclosure 1 6710 97-2345 Page 2 of 4 No.1820, Revision 0, hiay 1997, "Three hiile Island Nuclear Generating Station OTSG Kinetic Expansion Inspection Criteria Analysis". This report was presented at the July 25, 1997 GPU Nuclear /NRC meeting on this subject and was previously submitted to the NRC (GPU Nuclear letter to the NRC 6710 97-2348, dated August 8,1997). No changes are being made to plant stmetures, systems or components.

The Th11-1 OTSG tube inspections to be conducted in the upcoming 12R outage (September 1997) will utilize the inspection criteria and flaw dispositioning criteria described in the above referenced GPU Nuclear letter to the NRC as well as the criteria l submitted in TSCR #268 (6710-97-2357). Unacceptable defects will be repaired in accordance with the defined inspection criteria. The accident induced leakage postulated for the steam line break dose consequences calculation considers the potential sources of leakage from defects that remain in sersice.

Primary to-secondary leakage from postulated flaws is assumed to increase during a postulated steam line break accident. The kinetically expanded t 2be-to-tubesheet joint was originally qualified as leak-limiting and not leak tight. The total postulated primary-to-secondary leakage from defects that are dispositioned as acceptable will be limited to a value which will be bounded by the revised dose consequences described below. This will be accomplished by evaluating the possible tube leakage from each defect considering its surface extent, axial location within the tube, radial location of the tube in the generator, and steam line break tube load.

Accident analysis performed utilizing NRC Standard Review Plan (SRP) 15.1.5, Appendix A assumptions where applicable to the existing Th11-1 steam line break licensing basis shows that the environmental consequences of a bounding integrated primary-to-secondary accident induced leakage would result in an offsite dose of 29.7 rem thyroid and 0.12 rem whole body at the exclusion area boundary (EAB), and 8.5 rem thyroid and 0.02 rem whole body at the low population zone (LPZ), which is a small fraction of the 10 CFR 100 limits.

Utilizing Standard Review Plan (SRP) 15.1.5, Appendix / the revised steam line break dose consequences have been determined using the following conservative assumptions:

1. The unit has been operating with an RCS dose equivalent iodine activity of 1 pCi/g for the accident induced iodine spike analysis. This is the limit in Technical Specification 3.1.4. At the onset of the steam line break, the release rate ofiodine from the fuel rods to the RCS is assumed to spike by a factor of 500.

2. Some credit is taken for partitioning and radioactive decay afler the break.

3. The steam line break occurs between the Reactor Building and a turbine stop valve.

'4. Reactor coolant leakage into the steam generator continues until the Reactor Coolant System can be cooled down and the leakage terminated.

' Enclosure 1 6710 97-2345 Page 3 of 4 S. The total primary-to-secondary leakage assumption is integrated over the cooldown time to account for accident induced OTSG tube leakage.

NRC Standard Review Plan 15.1.5, Appendix A, " Radiological Consequences of htSL Failure Outside Containment of a PWR", identifies the applicable limit as "a small fraction of the 10CFR100 limit" (30 rem thyroid). The 10 "FR 50, Appendix A, GDC-19 limits for the control room are not affected by this change t'..ce the source term raumed for the TMI-l control room habitability analysis is based on the postulated Maximum Hypothetical Accident (MHA) which is bounding.

IV. Environmental Consideration GPU Nuclear has determined that this change to the TM1-1 Steam Line Break dose consequences involves no significant change in the amount or type of any effluent that may be released offsite, and that there is no significant increase in indNidual or cumulative occupational radiation exposure. The new radiological consequences of the revised steam line break dose conseqtmnces are below 10 CFR 100 limits for the EAB and LPZ. As such, I operation of TMI in accordance with the proposed change does not involve an unreviewed

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environmental safety question.

V. No Significant Hazards Considergian l

GPU Nuclear has <letermined that this License Amendment Request poses no significant hazards as dermea by 10 CFR 50.92.

1. Operation of the facility in accordance with the proposed amendment would nat involve .

a significant ir. crease in the probability of occurrence or the consequences of an accident l previously evaluated. This change has no effect on structures, systems or components i prior to the postulated steam line break accident or any other accident. OTSG tube l

. loads resulting from other postulated accidents are bounded by the calculated steam hne break accident tube loads. Other TMI-l design basis accidents, which could result in /

OTSG tube loads and environmental dose consequences, involve releases within the reactor building. These events generally result in rapid depressurization of the primary system which minimizes the differential pressure needed to establish a significant primary-to-secondary leak rate and the OTSG is isolated. Accordingly, leakage to the environment as a result ofinduced tube loads from postulated accidents other than steam line break in insignificant and therefore need not be considered. The existing  !

steam line break criteria is maintained in that OTSG structural integrity is assured and I p tulated doses remaia within 10 CFR 100 limits. The new radiological consequences  !

of the revised steam line break dose calculation are below 10 CFR 100 limits for the exclusion area boundary (EAB) and low population z%e (LPZ). % 10 CFR 50, i

Appendix A, GDC-19 limits for the control room are not affected by this change since '

the source term assumed for the TMI-l co. trol room habitability analysis remains bounding.

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3., Enclosure 1.

r .6710-97-2345-l- Page 4 of 4 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 previously evaluated. This change has no impact on any plant structures, systems or components. OTSG tube structural integrity is maintained. The only impact is the revised radiological o conseqt.ences of the steam line break analysis to account for hypothetical accident induced primary-to-secondary leakage.

3. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety. This change to the steam line break dose consequences does not involve a significant reduction in a margin of safety. The new radiological consequences of the resised steam line break dose calculation are below 10 CFR 100 limits for the EAB and LPZ, and do not affect the TM1-1 control room

> habitability analysis results. This change has no impact on any structures, systems or components.

VI. Implementation it is requested that the amendment authorizing this change become effective upon issuance..

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