ML20198E787

TS Change Request 266 to License DPR-50,incorporating Addl Sys Leakage Limits & Leak Test Requirements for Systems Outside Containment Which Were Not Previously Contained in TS 4.5.4

ML20198E787
Person / Time
Site:	Crane
Issue date:	07/30/1997
From:	Langenbach J GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20198E790	List: ML20198E787 ML20198E794
References
6710-97-2252, NUDOCS 9708080202
Download: ML20198E787 (15)

Text

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GPU Nucleer, Inc.

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Route 441 South NUCLEAR Post Othee Box 480

%ddletown, PA 17057 0480 Tel717 944 7621 July 30 1997 6710-97 h252 U.S. Nuclear Regulatory Commission Attention: Dc ument Control Desk Washington, DC 20555

Subject:

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

Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request (TSCR) No. 266 Accident Recirculation Systems Leakage Limits In accordance with 10 CFR 50.4 (b)(1), enclosed is TSCR No. 266 (Enclosure 1).

The purpose of this TSCR is to incorporate additional system leakage limits and leak test requirements for systems outside containment which were not previously contained in Technical Specification 4.5.4 nor considered in the TMI-l Updated FSAR (UFSAR) design buis accident (DBA) analysis dose calculations for 2568 MWt.

This TSCR also revises the Technical Specification 3.15.3 Bases for the Auxiliary and Fuel Handling Building Ventilation System (AFHBVS).

The revisions to Technical Specification 3.15.3 Bases for the AFHBVS serve to clarify system j

design requirements and accident analysis considerations. The revision states that the AFHBVS is l

not credited in reducing otT-site dose for the Maximum Ilypothetical Accident (MHA) or the Waste Gas Tank Rupture (WGTR) accident analysis dose calculations. The radiological consequences for the WGTR accident analysis are not increased by this TSCR.

Of Revisions to the TMI-l UFSAR design basis accident analyses descriptions of assumption (Enclosure 2) are being made to conform the UFSAR to the proposed Technical Specification

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changes. The revisions also address analytical discrepancies and non-conservatisms identified in NRC Inspection Report 96-201 and TMI-l License Event Report (LER)97-004, Revision 0. The revised radiological dose consequences for TMI-l Cycle 12 are based on the current licensed power level of 2568 MWt. The results of the MHA analysis have not resulted in an increase above the existing UFSAR values (recently reviewed by the NRC) for the exclusion area boundary (EAB).

In addition, the results continue to be below the 10 CFR 100 guideline limits for the EAB and the limits of General Design Criteria (GDC) 19 of 10 CFR 50 Appendix A for the control room.

9700000202 970730 l lll llljlfjll} l DR ADOCK 050 9

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U.S. Nuclear Regulatory Commission 6710-97 2252

' Page 2 The dose for the low population zone (LPZ) is increased slightly from the existing UFSAR values; however, the new values remain a small fraction of the 10 CFR 100 limits. The proposed UFSAR revisions will be included in the next revision of the UFSAR to be submitted following the next refueling outage, pursuant to 10 CFR 50.7l(e) as previously committed in the June 23, 1997

' response to IR 96-201.

Using the standards in 10 CFR 50.92, GPU Nuclear (GPUN) Inc. has concluded that these proposed changes do not constitute a signincant 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

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this request certifying sersice to the chief executives of the township and county in which the facility is lo:ated, as the designated official of the Commonwealth of Pennsylvania, Bureau of

- Radiation Protection.

Approval of this TSCR is requested as soon as possible in order that the TMI l Technical Specifications correctly reflect the post-accident recirculation leakage assumptions used in the calculation of MHA dose consequence analysis.

Sincerely, 4LL James W, Langenb ch Vice President and D1 ector, TMI' JWL/GMG/ lab

Enclosures:

(1) TMI-l TSCR No. 266 Safety Evaluation, No Significant Hazards Consideration, and Technical Specification Revised Pages (2) Proposed Revisions to TM1-1 UFSAR Pages (3) Certificate of Service for TMl-1 TSCR No. 266 cc:

Administrator Region I TMI Senior Resident inspector TMI Senior Project Manager

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. UNITED STATES OF AMERICA -

I NUCLEAR REGULATORY COMMISSION i

. IN TIIE hiATTER OF DOCKET NO. 50 289' i-GPU NUCLEARINC.

LICENSE NO. DPR 50 i

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CERTIFICATE OF SERVICE 4

1 This is to certify that a copy of Technical Specification Change Request No. 266 Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been 4

filed with executives of Londonderry Township, Dauphin County, Pennsylvania; Dauphin County, Pennsylvania; and the Pennsylvania Depanment of Environmental Resources, Bureau of Radiation Protection, by deposit in the United States mail, addressed as follows:

Mr. Darryl LeHew, Chairman Ms. Sally S. Klein, Chairman _

Board of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County-R. D. #1, Geyers Church Road Dauphin County Coudhouse Middletown, PA '17057 Harrisburg, PA 17120 Director, Bureau of Radiation Protection PA Dept. ofEnvironmental Resources Rachael Carson State Office Building P.O. Box 8469

- Harrisburg, PA 17105-8469 Att: Mr. Stan Maingi GPU NUCLEAR INC.

BY: 9W gice President and Diregr, Tui 7f3D !97 DATE-I

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h1ETROPOLITAN EDISON COh1PANY JERSEY CENTRAL POWER & LIGliT COh1PANY AND

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PENNSYLVANI A ELECfRIC COhtPANY l-TilREE hilLE ISLAND NUCLEAR STATION, UNIT 1 3

Operating License No. DPR-50 l

Docket No. 50-289 l

Technical Specification Change Request (TSCR)No 266 i

l COhih10NWEALTH OF PENNSYLVANIA

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l COUNTY OF DAUP111N

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l This TSCR is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three hiile Island Nuclear Station, Unit 1. As a part of this request, j

proposed replacement pages for the Appendix A Technical Specifications are also included. All statements contained in this submittal have been reviewed, and all such statements made and matters set forth therein are tme and correct to the best of my knowledge.

GPU NUCLEAR INC.

BY:

M WA yce President and D ttor, Thil Swom a Subsc '

to before me thisi ayo 1997.

Le2du

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/ Notary Public

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Nolanal Seal l

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Su: anne C. M*tesik, Notary Public Londonderry Twp., Dauphin County Lty Commusion Expires Nov 22,1999 Member, PsecHvan e skem?mn Of Nnntes

ENCLOSURE 1 TMI-l TSCR No. 266 Safety Evaluation,_

No Significant Hazards Consideration and 1

Proposed Technical Specification Revised Pages i

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S 6710-97-2252 Page1of9 1.

IECHNICAL SPECIFICATION ClI ANGE REQUEST (TSCR) NO. 266 GPU Nuclear (GPUN) Inc. requests that the following changed replacement pages be inserted into the existing Technical Specifications :

Revised Technical Specification Pages: iii, 3-62d,4-45 These pages are attached to this Enclosure.

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RI3ASON FOR CHANGli The purpose of this TSCR is to incorporate additional system limits and leak test requirements in Technical Specification Section 4.5A, for systems outside containment.

These leakage criteria were not previously contained in the Technical Specifications, nor considered in the TM1-1 UFSAR dose calculations for the Maximum Hypothetical Accident (MHA).

Technical Specification Section 4.5.4 is re-titled from: " Decay lleat Removal System Leakage," to: " Accident Recirculation Systems Leakage" based on changes to this section which add leakage and leak testing criteria for the Building Spray (BS) and Make-Up (MU)

Systems The total allowable leakage is changed from 6 gph (0.1 gpm) to 18 geh (0.30 gpm) to be consistent with the assumption used in the revised MHA analysis described below. The Technical Specification Section 4.5.4 Bases is also revised to ref1cet these proposed changes and to remove the specific calculated dose values since these are to be included in the referenced UFSAR Sections.

The Tab!c of Contents page iii is revised to reflect the new title of Technical Specification Section 4.5.4.

This TSCR also revises the Technical Specification Section 3.15.3 Bases for the Auxiliary and Fuel Handling Building Air Treatment System ('a.k.a., the Auxiliary and Fuel Handling Building Ventilation System - AFHBVS) to clarify system design requirements and accident analysis considerations. The revision serves to state that the AFHBVS is not credited in reducing off-site dose for the MHA or the Waste Gas Tank Rupture (WGTR) accident dose calculations. However, the AFHBVS could be used to reduce off-site dose if the system is available. This Bases clarification also identifies that the Fuel Handling Building ESF Air Treatment System is utilized for the Fuel Handling Accident in the Fuel !Iandling building consistent with Technical Specification 3.15.4.

The References section of Tecimical Specification 3.15,3 Bases is also updated to include appropriate UFS AR sections.

6710-97-2252 Page 2 of 9 Thers proposed changes to modify the Thil 1 Technical Specifications are being made pursuant to the GPUN connaitment and schedule defined in our response letter to IR 96-201, dated June 23,1997 (GPUN Letter No. 6710-97-2242). A revised MilA dose consequence analysis (Calc. C-1101 202-E260-329) was performed to ir Jedx 1.

The effects ofincreased emergency core cooling system (ECCS) leakage into the Auxiliary Building, without taking credit for AFIIBVS charcoal filters. This revises the total assumed leak rate from 6 gph (0.1 gpm) to 18 gph (0.3 gpm). This change in assumed leakage rate also eliminates the discrepancy between the Technical Specification limit (6 gph) and the UFSAR assumptio w ; gph) described for the hillA, identified in NRC Inspection Report 96-201 dateo april 15,1997 and Th11 1 Licensee Event Report (LER)97-004, Revision 0, dated April 4,1997. The new leak rate includes consideration of the Make-Up and Building Spray systems contributions to the assumed total leakage rate.

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The efTects of leakage through the ECCS boundary valves to tanks vented to atmosphere (3 gpm) not previously considered.

3, The rate of mixing within the Reactor Building between the sprayed and unsprayed volumes was increased from 54,000 cfm to 00,000 cfm.

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An updated X/Q from 8.3 x 10 sec/m'to 6.8 x 10~'sec/m, as previously reviewed d

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A change in containment leak rate from 0.12% per day to 0.1% per day based on the TM1-1 Technical Specification limit. The 0.1% limit was previously used for the Cycle 10 dose consequence analysis, as documented in the current UFSAR Appendix 14C.

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A Flashing Fraction decreased from 4.0 % to 1.25 %, based on post-accident sump temperature, in addition, the UFSAR description is being revised to clearly identify that the AFliBVS charcoal filters are not credited in dose calculations for the Waste Gas Tank Rupture (WGTR).

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6710-97-2252 Page 3 of 9 111.

SAFETY EVALUATION JUSTIFYING CHANGE The proposed TSCR incorporates changes to Technical Specification Sections 3,15.3 and 4.5.4 to reficct the revised accident analysis assumptions. This change request does not require any modifications to plant stmetures, systems, or components. The new radiological consequences of the revised MllA dose calculation are below the existing UFSAR values for the EAB and below the 10 CFR 100 guideline limits. The dose for the low population zone (LPZ) is increased slightly from the existing UFSAR values however, the new values remain a small fraction of the 10 CFR 100 guideline limits. In addition, the Control Room dose remains below the 10 CFR 50, Appendix A, GDC-19 limits. This change has no effect on the loss-of-coolant accident (LOCA) safety analpis for emergency core cooling system (ECCS) performance, which demonstrates conformance to the acceptance criteria of 10 CFR 50.46, as described in the TMI-l UFS AR Sections 6.1,14.2.2.3, and 14.2.2.4 The description of the WGTR accident in the TMI I UFS AR Section 14.2.2.6 states that the reactor coolant passes through purification demineralizers which remove 99% of the iodine, d

and implies that charcoal filters in the Auxiliary Building with 90% efliciency would i

remove more iodine as the radioactivity releases into the environment, llowever, the calculation of the radiological consequences for this accident did not take credit for lodine i

removal via the purification demineralizers or the AFIIBVS charcoal filters. The increased ECCS leakage does not afTect this accident. Thus, the Technical Specification Section 3.15.3 Bases and description of the WGTR accident in UFSAR Section 14.2.2.6 is being revised to clarify that no credit was taken.

The MHA postulates a gross release of fission products to the reactor building. The release is not mechanistic and a specific means for it to occur is not postulated. The purpose of the evaluation is to determine if the oiTsite dose consequences are acceptable. This accident was re-analyzed accounting for the following revised assumptions:

1. ECCS leakage through mechanical jo:nts into the Auxiliary Building of 0.3 gpm, as reflected in the proposed TMI l Technical Specification (TS) Section 4.5.4 revision.

The 0.3 gpm ECCS leakage acceptance criteria for TS 4.5.4 was established based on the results of system leakage inspections throughout plant history and the acceptability of the resultant MHA dose consequence with the proposed acceptance criteria. The previous TS limit of 0.1 gpm was for DH system leakage only and the new Technical Specification limits are increased to account for leakage from two additional systems, Building Spray and Make-Up.

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2. No credit for the AFHBVS, as reflected in the proposed TMI-l Technical Specification Section 3.15.3 Bases clarification. On February 28,1997, GPUN declared the AFliBVS inoperable due to quality classificatiot. concerns. The duct and fan system is commercial grade and is not powered from the emergency diesel generators; however, the system was credited in the Technical Specification Bases for mitigating dose rates for DBAs to below 10 CFR 100 limits. Since the AFiiBVS is not safety grade pursuant to Regulatory Guide 1.52, the Bases in the Technical Specification 3.15.3 is being revised to remove credit for dose mitigation.

3. ECCS leakage through boundary valves to tanks vented to atmosphere of 3 gpm.

The 3 gpm acceptance criteria for ECCS boundary valve leakage to tanks vented to atmosphere is based on the resolution capability of new leakage detection tests and the acceptability of the resultant dose consequence of the MHA analysis with this proposed criteria, which were not previously considered as re.' ease paths.

Other conservatisms and assumptions are discussed below, and are consistent with those used in the existing UFSAR analysis, or previously reviewed by NRC.

The magnitude of the radioactive release is based on the fission product buildup from fuel burnup. The isotopic core inventory was based on the current UFSAR power level of 2568 MWt. One hundred percent of the noble gases and fifty percent of the lodine (s) in the core are assumed to be released into the reactor building atmosphere. Funber, only 50% of the lodine (s) released to the reactor building are assumed to plate out.

Decay of fission products are assumed to occur while they are confined to the Reactor Building, but are not assumed to occur once they pass to the environment. Reactor building pressure is assumed to be at design pressure; therefore, reactor building leakage is assumed to be at its design leak rate for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The reactor building leak rate is then assumed to be at one halfits design value for the next 29 days. lodine removal from the reactor building atmosphere is performed assuming one spray pump header and one cooling fan. The resulting 2-hour Exclusion Area Boundary (EAB) dose is 189 rem to the thyroid, and 6.0 rem to the whole body. The corresponding 30 day Low Population Zone (LPZ) Boundary dose is 13.0 rem to the thyroid, and less than I rem to the whole body. These off-site doses, which were calculated on the basis of all the assumptions above and conservatively rounded upward to the next whole integer, remain below the 10 CFR 100 guideline limits. This rounding upward to the next whole integer also provides additional margin for slight variations in future fuel cycles. These results funher serve to confirm that the classification of the AFHBVS as a non Safety-Related system is appropriate, and that the system is considered to be operable in accordance with the Technical Specification 3.15.3 requirements.

6710.?"' 2252 Page5of9 The 30-day control room habitability doses were also evaluated by incorporating the additional MHA release paths. The principal assumptions and methods approved by the NRC remain applicable to this new evaluation. The revised Control Room doses are 0.83 rem to the whole body and 14 rem to the skin, which are below 10 CFR 50, Appendix A, GDC 19 limits. As concurred by the NRC (NRC letter dated August 14, 1986, "TMI l Control Habitability Review" Supplemental Safety Evaluation Report),

only beta skin and gamma whole body dose is addressed, pending resolution of the lodine source term.

Additionally, the MilA analysis parameters are being revised in the enclosed UFSAR description which reflect the latest values, as follows:

1) Mixing Flow Between Sprayed and Unsprayed Volumes A two-compartment model is used to calculate spray iodine removal in the containment since certain areas are not reached by droplets. It is assumed that only one out of two containment spray trains and only one out of three reactor building emergency cooling units (fans) are in operation, consistent with single-failure assumptions. The volumetric flow rate between the sprayed and unsprayed areas was evaluated using a three-dimensional model of the containment. The average exchange rate between the sprayed and unsprayed areas in the reactor building over a range of post-LOCA reactor building conditions was determined to be at least 100,000 cfm with one fan and one spray pump in operation, as compared with 54,000 cfm used in the existing UFSAR description.

2) Average Atmospheric DitTusion Factor The atmospheric difTusion factor X/Q, at the exclusion boundary for a two hour ground level release used in the existing UFSAR MHA analysis was 8.3 x 10d 3

sec/m. Subsequent evaluations of short term (0-2 Hrs) accident meteorology using a two year period of onsite meteorological data with better than 90% data recovery, 4

established an accident X/Q of 6.8 x 10 sec/m, as previously submitted to NRC in GPU letter dated May 8,1979 (GQL-0460).

3) Containment Leakage Rate The existing UFS AR MHA analysis used a containment leakage rate of 0.12 %/ day.

However, TMI-l Technical Specifications restrict the containment leakage rate to 0.1 %/ day, this value is used in the revised MHA dose calculation. The dose calculation for Cycle 10 used the Technical Specification limit as documented in GPU Nuclear letter to the NRC data June 7,1993 (C311-93-2070) in support of TMI-l License Amendment No.178, dated September 10, 1993. This Technical Specification leakage rate of 0.1 %/ day and its use in the Cycle 10 analysis is reflected in the existing TMI-l UFSAR, Appendix 14C.

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6710-97 2252 Page 6 of 9

4) Flashing Fraction Ti ? current UFSAR, Table 14.2 20, identifies a liquid leakage rate of 2,165 ml/hr ana 90 ml/hr leakage that flashes to steam. This is the equivalent of approximately 4% flashing, and appears to be based on a sump temperature of 250 F. The revised dose analysis is based on the sump temperature profile during post-accident -

recirculation. The sump temperature decreases below 224 F in about two minutes afler the stan of recirculation. The assumption of a constant sump temperature of 224 F and a corresponding flashing fraction of 1.25% is conservative for the two hour post-accident period.

Therefore, it is concluded that the propmed changes to the Technical Specification Sections 3.15.3 and 4.5.4 do not adversely affect nuclear safety or safe plant operations. The revised dose consequences were calculated using the NRC approved methodology contained in the UFSAR Appendix 14C. The new radiological dose consequences of the revised MHA analysis are below the values identified in the UFSAR for the EAB and below the 10 CFR 100 guideline limits for both the EAB and LPZ, and below the 10 CFR 50, Appendix A, GDC-19 limits for the control room.

Environmental Considera1!nD GPUN has determined that this change to the TM1-1 UFSAR Technical Specification Sections 3.15.3 and 4.5.4 involves no significant change in the amount or type of any efiluent that may be released off-site, and that there is no significant increase in individual or cumulative occupational radiation exposure.

The new radiological consequences of the revised MHA dose consequences are below the 10 CFR 100 guideline limits for the EAB and LPZ.

6710-97-2252 Page 7 of 9 TABLEI COMPARISON OF POST ACCIDENT MilA DOSES TIIYROID Arca/E

.! UFShR Analysis:-

(Revised Analysis 210 CFR 100.11r Coatsuter--

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(rem)

- (mn)'-

Exclusion Area Boundary 189 189 300 Total (2-hour consequence)

Low Population Zone 8.8 13.0 300 (30-day consequence)

WilOLE HODY

. l? Area /:.

UFSAR Analysis:

Revised Analysis?

(10 CFR 100,11(

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(rem)4 Exclusion Area Boundary 7.6 6.0 25 Total (2-hour consequence)

Low Population Zone 0.21

<1 25 (30-day consequence)

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iInclosure 1 6710-97 2252 Page 8 of 9 11

@ SIGNIFICANT llAZARDS CONSIDERATION GPUN has determined that this TSCR poses no significant hazards consideration as defined by 10 CFR 50.92.

1. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability of occurrence or the consequences of an accident previously evaluated.

No physical modifications which would change structures, systems, or components are being made or proposed by this TSCR. This change has no afTect on the LOCA safety analysis for ECCS performance. The results of revised MilA dose calculation are less than that previously evaluated in the UFSAR for the exclusion area boundarf (EAB). In addition the doses are below the 10 CFR 100 guideline limits for both the EAB and low population zone (LPZ) as shown in Table 1, and below the 10 CFR 50 Appendix A, GDC-19 limits for the control room. The LPZ increases in dose consequence are the result of using more conservative assumptions in the revised analyses and the new values remain a small fraction of the 10 CFR 100 !imits The WGTR dose calculation is not affected by this TSCR.

The proposed Technical Specification changes ensure that the MilA and WGTR accident analysis parameters remain bounded during plant operation.

2 Operation of the facility in accordance with the pioposed amendment would not create the possibility of a new or different kind of accident from any previously evaluated. This TSCR does not involve any physical modifications which would affect stmetures, systems, or components, nor does it involve any changes in plant operation. The only changes resulting from this TSCR are revisions to leakage limits and testing requirements necessary to reflect the revised MilA analysis and to correct discrepancies identified by the NRC, as referenced above. Therefore, the proposed change does 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. This TSCR doet not involve changes to Technical Specification defined Safety Limits, Limiting Conditions for Operation, and does not involve any change to safety system setpoints for operation. Therefore, the proposed change does not involve a significant reduction in the margin of safety.

6710 97-2252 Page 9 of 9 111.

IMPLEMENTATION It is requested that the license amendment authorizing this change become efTective upon issuance.

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Proposed Technical Specification Revised Pages s

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