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Application for Amend to License DPR-50,deleting Tech Spec Requirements in Section 4.1.3,Table 4.1-4,Item 2 for Reactor Bldg High Range Area Gamma Monitor
	ML20064B485
Person / Time
Site:	Three Mile Island
Issue date:	10/04/1990
From:	Hukill H; GENERAL PUBLIC UTILITIES CORP.
To:	;
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ML20064B463	List: ML20064B462; ML20064B464; ML20064B485;
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	NUDOCS 9010170087
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T-METROPOLITAN EDISON COMPANY r

JERSEY CENTRAL POWER & LIGHT COMPANY AND PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION, UNIT 1 ,

t 4-Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No.195 w

This Technical Specification Change Request is submitted in support of Licensee's '

. request to change Appendix A to Operating License No. DPR-50 for'Three Mile Island

., Nuclear Station, Unit 1. As a part of this request, proposed replacement pages t for Appendix A' are also included.

GPU NUCLEAR CORPORATION l

6 BY: . -

Vice Pre'sYdent & Director, TMI-1 Sworn and subscribed 'g to beforg mp this '

- day of 6' eft bR. ,1990.

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x 1LAL Y tAMGM Notary Public L meswsw ,

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I UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 1

IN THE MATTER OF DOCKET NO. 50-2B9 GPU NUCLEAR CORPORATION LICENSE NO. DPR-50 CERTIFICATE OF SERVICE This is to ce'tify that a copy of Technical Specification Change Request No. 195 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with executives of Londonderry Township, Dauphin County, Pennsylvania; Dauphin County, Pennsylvania

and the PennsylvaniaDepartme0tofEnvironmentalResources,BureauofRadiatIon Protection, by deposit in the United States mail, addressed as follows: ,

l Mr. Kenneth E. Witmer, Chairman Ms. Sally S. Klein, Chairman I Board of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County 25 Roslyn Road Dauphin County Courthouse Elizabethtown, PA 17022 Harrisburg, PA 17120 Mr. Thomas Gerusky, Director PA. Dept. of Environmental Resources !

Bureau-of Radiation Protection l l P.O. Box 2063 Harrisburg, PA '17120 GPU NUCLEAR CORPORATION '

', BY: , < 1 Vice Prdsident & Director, TMI-1 l l

DATE: October 4. 1990 ;

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I. TECHNICAL SPECIFICATION CHANGE REOUEST NO. 195 GPUN requests that the attached revised pages replace 3-40b, 4-Sa and 4-105

, of the TMI-1 Technical Specifications.

II. Reasons for the Chances This change is being submitted to revise two Technical Specifications sections and modify the bases statements supporting an unrelated Technical Specification as described below.

U The Technical Specification changes involves i

1. page 4-Sa - deletion of the Technical Specification requirements in F section 4.1.3, Table 4.1-4, Item 2, for the Reactor Building High Range Area Gamma Monitor RM-G8, and

2. page 4-105 - revision of the notation in Table 4.22-2, j. to include an action to prepare a report within 30 days of discovery of a condition resulting in the collection of a nonrepresentative sample at the condenser Vent Stack Continuous Iodine Sampler. This action ,

statement would apply when the sampler or alternate sampling equipment is not placed in service within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

The bases statenent change involves page 3-40b -

clarification of the bases statements supporting Technical Specification Table 3.5-3 to provide distinction between the ,

design basis and beyond design basis accident monitoring instrumenta-

. tion. ,

III. Safety Evaluation Justifyino Technical Specification Chance 1 Change 1 would delete radiation monitor RM-G8 from the Technical Specifica-tions and rely instead on RM-G22 and RM-G23. This change is supported by the following justification.

The Reactor Building High Range Area Gamma Radiation Monitor, RM-GB, is a l containment dome monitor which is part of the original plant configuration.

It is a high range gamma sensitive ionization chamber detector capable of measuring radiation levels from 0.01 R/hr to 106 R/hr. Surveillance 1

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l requirements for RM-G8 are specified in Technical Specification 4.1-1, j item 28. 5 While the reactor is in operation, monitor RM-G8 is not relied upon for j personnel protection. Monitor RM-G8 is not interlocked with any plant protective features. Monitor RM-G8 is equipped with a high radiation level alarm trip set point corresponding to conditions indicative of a loss of coolant accident.

Monitors RM-G22 and RM-G23 which have a detection range from 1 to 10 7 R/Hr j

were installed to satisfy NUREG 0737 and Regulatory Guide 1.97 require-ments. These monitors are also provided with alarm set points correspond-ing to the conditions indicative of a LOCA. The detectors are constructed with ion chambers which are strong enough to withstand LOCA pressures but thin enough to provide full response to low energy photons. Monitors RM-G22 and RM-G23 are designed for continuous operation during all plant operating modes and abnormal conditions, operability and surveillance requirements for these monitors are specified in Technical Specification 3.5.5 and 4.1.3, Table 4.1-4 Item 2 respectively.

RM-G8 provides input to the Safety Parameter Display System. That input to SPDS will be replaced by input from RM-G22 and RM-G23, enhancing the capability of SPDS. The Safety Parameter Display System is beyond the scope of Technical Specification requirements.

It is CPUN's intention, on approval of this TSCR, to electrically isolate, lif t and spare the leads to RM-G8 and abandon it in place until its removal at a later date. RM-G22 and RM-G23 will more reliably perform post-accident containment monitoring.

IV. No Sionificant Hazards Consideration for Chance 1 GPUN has determined that this Technical Specification change request poses no significant hazards as defined by 10 CFR 50.92 and operation of the f acility in accordance with the proposed amendment will have no adverse effect on nuclear safety or safe plant operations.

Monitor RM-G8 was not designed to support normal plant operation. RM-G8 is a high range radiation monitor lo'ated in the containment dome which provides indication of post-accident radjation levels. Removal of RM-G8 has no effect on the probability of occurrecce of an accident previously evaluated.

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RM-G8 was intended to provide information to assist in following the co.rse of a design basis accident. RM-08 has no tratomatic interlock featutes.

Monitors RM-022 and RM-G23 were installed in response to NUGO 0737 to provide diagnostic information regarding high radiation lev".,is in contain-ment following a design basis accident. Monitors RM-G2* and RM-G23 satisfy Regulatory Guide 1.97 requirements, exceed the post cacident capabilities provided by RM-08 and are considered suitable replacements for RM-G8.

Technical specifications 3.5.5 and 4.1.3 ensure the functional capability of these monitors. Therefore removal of RM-G8 from the plant and from the Technical specifications has no impact on the consequences of an accident previously evaluated.

As discussed above, use of the proposed monitoring would not create the possibility of a new or different type of accident from any accident previously evaluated. Neither will use of the proposed monitors involve

c. significant reduction in a margin of safety. The margin of safety for the proposed monitors is no less than that afforded by the existing Technical specifications.

V. Safety Evaluation Justifyino Technical Soecification Chance 2 Table 4.22-2, note j, does not presently provide a practical action to be taken if the surveillance requirement is not met. The notation, as it exists, is inconsistent with similar tabular notations in that it does not require action to be taken if the surveillance requirements for obtaining a representative sample and performing analyses in accordance with the sampling and analysis program are not met.

This proposed Technical specification change adds an action requirement if a condition resulting in a nonrepresentative sample is discovered and the sampling equipment (original / alternate) is not put in service within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . A report will be prepared and submitted within 30' days describing (a) the cause of the inoperability; (b) the action (s) taken to restore representative sampling capability; (c) the action (s) being taken to prevent recurrence; and (d) quantification of releases via this pathway during the period of inoperability and a comparison to the limits prescribed by Technical specification 3.22.2.1.b.

The action defined by this Technical specification change provides additional control to demonstrate that the Limiting Condition for

-operation in Technical specification 3.22.2.1.b is not exceeded when the normal surveillance requirement cannot be met.

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VI. No Sionificant Hazards Consideration for Chance 2 i

By limiting the dose rate due to radioactive materials in gaseous ef fluents released from the site, exclusive of noble gases, to less than or equal to 1,500 mrom/ year to any organ, the existing Technical Specificatione ensure protection of the public health and safety.

i A condenser vent iodine sampling capability is provided to ensure that releases of radiciodine through this effluent pathway during normal plant f operations result in no more than a fraction of the annual dose limits specified in 10 CFR 20 for individuals in unrestricted areas. This off line sampling capability provides a means for quantifying ef fluent releases through this path. No aspect of this sampler or sampling capability represents an accident precursor, or af f ects in any way the probability of occurrence of any accident previously evaluated. )

This sampler and sampling capability are not relied upon to perform a post-accident function. Thus, no aspect of this sampler or sampling capability can have an impact on the consequences of any accident previously evaluat- .

ed. Application of the proposed revised surveillance action requirement presents no increase in the probability of occurrence or the consequences of any accident previously evaluated.

Because the sampler and sampling capability are independent of nuclear plant operation, the proposed revised surveillance action requirement does not increase the probability of occurrence or the consequences of_an accident or malfunction previous 1'; evtluated. A new or different type of accident or malfunction is not created which differs from any previously evaluated.

l The proposed revised surveillance requirement does not reduce the margin of safety afforded by this Tectnical Specification limit; rather, it pro-l vides an additional administrative control to ensure and demonstrate that i

this limit is not exceeded. Use of the revised surveillance action

l. requirement does not involve a reduction in a margin of safety.

Bases Statement chance A is suooorted by the followino:

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The existing statement identifies the containment water level and pressure instrument and various high range radiation monitors as useful for beyond design basis accident parameter monitoring.

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i Those instruments are for design basis events. When included with the containment hydrogen concentration monitor, they are useful as a group for evaluating and predicting the course of beyond design basis accidents. The wording as revised clarifies the function of the instruments. j I

This editorial change does not involve plant configuration modifications, )

set point alterations, or reduction in established safety limits. l l

VII. Imolementation i

It is requested that the amendment authorizing these changes become effective on issuance. It shall be implemented within thirty days of receipt.

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ML20064B485

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