ML20127P720

Proposed Tech Specs Consisting of Change Request Number 194 Suppl 1,revising TS 6.9.4,to Change Reporting Requirements from Semiannual to Annual, Per 10CFR50.36(a)(2) Effective 921001

ML20127P720
Person / Time
Site:	Three Mile Island
Issue date:	11/30/1992
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20127P711	List: ML20127P707 ML20127P720
References
NUDOCS 9212030084
Download: ML20127P720 (12)

Text

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.. Three Mile Island Nuclear Plant, Unit 1 (TM1-1)

. Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No.194, Supplem nt 1 1.0. SUPPLEMENT l TO TECHNLCAL_SflCIFICATION CHANGE RE0 VEST (TSCR) NO.191 GPUN requests that the following pages of the THI-l TSCR No. 194 be 1 replaced, as indico nd below:

Replace Pagan v.; l-6; 6-18; 6-23; and, 6-24.

GPUN requests that the following amended pages of the THI-1 Technical Specifications (Tech. Specs.) be added to TSCR No. 194, as indicated below:

Add Paaes: 3-96; 3-100; 4-99; and 4-105.

2.0. RESCRIPTION OF SUPPLEMENTAL CHANGES The following pages have been changed by deletion of the word

" semiannual" in reference to the periodic reporting of effluent releases, and replacement with the word " annual" pursuant to 10 CFR 50.36(4)(2).

1. pv This Table of Contents page has been updated to reflect the administrative correction necessary for the fech. Spec. 6.9.4 title change.

2. p 1-6 Tech. Spec. Definition 1.15, "Offsite Dose Calculation Mancal l (0DCM)" conforms to 10 CFR 50.36(a)(2).

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3. p 3-96 Tech. Spec. 3.21.1(b), " Radioactive Liquid Effluent Instrumentation" LCO reporting requirement conforms to 10 CFR 50.36(a)(2).

,. 4. p 3-100 Tech. Spec. 3.21.2(b), " Radioactive Liquid Effluent L Instrumentation" LC0 ryorting requirement conforms to 10 CFR 50.36(a)(2).

5. p 4-99 Tech. Spec. 4.22.1, Table Notations (f.) for Table 4.27.-1,

" Radioactive Liquid Waste Sampling and AnalvC s Program"~

conforms to 10 CFR 50.36(a)(2).

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6. p 4-105 Tech. Spec. 4.22.2, Table Not ions (9.) for Table 4.22-2,

" Radioactive Liquid Waste Sampling and Analysis Prograt" j conforms to 10 CFR 50.36(a)(2).

7. p 6-18 Tech. Spec. 6.9.4, " Radioactive Effluent Release Report-" has been revised to reflect annual reporting rather than the L previous " semiannual" reporting of releases made pursuant to i Appendix I to 10 CFR Part 50.

8. p 6-23 Tech, Spec. 6.13, " Process Control Program (PCP)" has been updated to conform to 10 CFI 50.36(a)(2).

l 9. p 6-24 Tech. Spec. 6.14.2, "Offsite Dose Calculation Manual (00CM)"

9212030084 9211. , ~~ 7 9 updated to conform to 10 CFR 50.36(a)(2).

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Three Hi.e Island Nuclear Plant, Unit 1 (THI-1)

Incftnical Sogcification Chanae Reouest Ndi4. Supplement 1 ,

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'3 0. . BEASONS ffft.1UPPLEMENTAL CHANGE 3.1. Administrative changes only to conform the Technical Specificaticas to the current rules and regulations pertaining to the reporting requirements of 10 CFP. 50.36(a)(2).

4.0 SAFETY EVALUATION JUSTIFYING CHANGE 4.1 Not appliraole. Administrative changes only. See also TSCR-194.

Health and Safety of the public is not decreased by these changes made to conform the Technical Specifications to the revi;ed rules under 10 CFR 50.36(a)(2).

5.0 N0 SIGNIFICANT HAZARDS CONSIDFRATIONS GPUN iias determined that this Technical Specification Change Request as supplems )ed herein, involves nc significant hazards consideration as defined iy NRC in 10 CFR 50.92.

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

The propored cnanges to TSCR 134 allow reporting of effluent releases pursuant to 10 CFR 50.36(a)(2) on an annual basis rather than semi-annually. This proposal simplifies the radiological effluent Toch. Specs.(RETS), meets the regulatory requirements for radioactive effluent and radiological environmental monitoring, and is considered a line-item improvement of the Tech. Specs. Thus, this change does not increase the probability of occurrence or consequences of cc accident previously evaluated.

5.2 Operation of the facility in accordance with the proposed supplement would not create the possibility of a new or different kind of accident from any accident previousiy evaluated.

The proposed changes to TSCR 194 allow reporting of effluent releases pursuant to 10 CFR 50.36(a)(2) on an annual basis rather than semi-annually. Future changes to procedural detaH s in the ODCM and the PCP will be handled under the administrative. controls for changes to these documents. Therefore, this change has no ai'fect on the possibility of creating a new or different kind of accident- from any accident previously evaluated.

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.- Three Mlle Island Nuclear Picat, Unit 1 (TMI-1)-

lechnical Specifica* ion Chanae Reouest No. 194. Suorlement 1 5.3 Operation of the facility in accordance with the proposed supplement would not involve a significant reduction ir, a margin of safety.

The proposed changes to TSCR 194 allow reporting of effluent releases pursuant to 10 CFR SG.36(a)(2) on an annual basis rather than semi-a.nnually. Therefore, it is concluded that operation of the facility in accordance with the proposed amendment does not involve a significant reduction in a margin of safety.

The Commission has provided guidelines on the 9pplication of the ,

three standards by listing specific examples in 45 FR 14870. The proposed changes are considered to be in the same category as example (vi) of amendments that are considered not likely to involve significant hazards consideration in that the proposed change conforms the Tech. Specs. ta changes made to 10 CFR 50,36(a)(2).

Implementation of the propored changes to TSCR 194 &ccording to 10 CFR 50.36(a)(2) conforms the Technical Specification reporting requirements b NRC rules and regulation. Thus, operation af th facility in accordance with the proposed changes involves no significant hazards considerations.

6.0 IMPLEMENTATION No change from TSCR 194

7.0 REFERENCES

No change from TSCR 194 i

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.. TABLE OF CONTENTS Section bag 5- DESIGN FEATURES 5-1 5.1 11T[ 5-1 5,2 CONTAINMENT' 5-2 5.2.1 REACTOR BUILDING 5-2 5.2.2 REACTOR BUILDING ISOLATION SYSTEM 5-3 '

5.3 REACTOR 5-4 5.3.1 REACTOR CORE 3-4 5.3.2 REACTOR COOLANT SYSTEM 5-4 5.4 NEW AND SPENT FUEL STORAGE FACfLITIES 5-6 5.4.1 NEW FUEL STORACE 5-6 5.4.2 SPENT FUEL STORAGE 5-6 5.5 AIR INTAKE TUNNEL FIRE PROTECTION SYSTEMS 5-8 6 ADMINISTRATIVE CONTROLS 6-1 6.1 RESPONSIBILITY 6-1 ,

6.2 ORGANIZAljQN 6-1 6.2.1 CORPORATE 6-1 5.2.2 UNIT STAFF 6-1 6.3 UNIT STAFF OVALIFICATIONS 6-3 6.4 TRAINING 6-3 6.5 REVIEW AND AUDIT 6-3 6.5.1 TECHNICAL REVIEW AND CONTROL 6-4

6. 5. 2 - INDEPENDENT SAFETY REVIEW 6-5 6.5.3 AUDITS 6-7 6.5.4 INDEPENDENT ONSITE SAFETY REVIEW GROUP 6-8 6.6 REPORTABLE EVENT ACTION 6-In 6.7 SAFETY LJMIT VIOLATION 6-10 6.8 EROCEDURES AND PROGRAMS 6-11 6.9 REPORTING 9E0VIREMENTS 6-12 ,

6.9.1 ROUTINE REPORTS 6-12 6.9.2 DELETED 6-14 S.9.3 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6-17 6.9.4 ANNUAL PADI0 ACTIVE EFFLUENT RELEASE REPORT 6-18 l 6.9.5 CORE OPERATING LIMITS REPORT 6-19 6.10 REC!lRD RETENTION 6-20 6.11 RADIATION PROTECTION PROGRAM 6-22 6.12 HIGH RADIATION AREA 6-22 6.13 PROCESS CONTROL PROGRAM 6-23 6.'14 0FFSITE DOSE CALCULATION MANUAL (00CM) 6-24 6.15 DELETED 6-24 6.16 POST ACCIDENT SAMPLING PROGRAMS 6-24 NUREG 0737 (II.B.3. II.F.1.2) 6.17 MA)2R CHANGES TO _R6DJ0 ACTIVE WAITf , TREATMENT SY77 EMS 6-25

-v- j Amendment No. JJ, 97,72,77,127,150 l l

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1.9 DELETED 1.1.0 DELETED l.11 DELETED ,

1.12 DOSE E0lllVALENT l-131 The DOSE E0VIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, I-133, I-134 and :-135 actually present. The thyroid dose conversion factors usad for this

,- calculation shall be those listed in Table'lII of TID 14844, " Calculation of Distance F.ctors for Power and Test Reactor Sites". [0r in Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977.]

1.13 SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel res5 case when the channel sensor is exposed to a radioactive source.

1.14 DELETED 1.15 '0FFSITE DOSE CALCULATION MANUAL (ODCM)

The OFFSITE DOSE CALCULATION MANUAL (00CM) shall con +ain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluent, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Satpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radinactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Onerating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.3 and 6.9.4.

1.16 PROCESS CONTROL PROGRAM (PCP)

The PROCESS CONTROL PROGRAM (PCP) shall contain the current formule.s, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CF2 Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

l 1.17 GASEOUS RADWASTE TREATMENT The GASE0US RADWASTE TREATMENT SYSTEM is the system designed and installed to reduce radioactive gaseous effluent by collecting primary coolant = system of f gases from the primary system and providing for. delay or holdup for the purpose of raducing the total radioactivity prior to release to the envi ron.,ient .

1-6 Amendment No. 77, 137 i

3.21 RADI0 ACTIVE EFFLUENT INSTRUMENTATIQN 112J21 RADI0 ACTIVE L10VID EFFLUENT INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.21.1 The radioactive liquid effluent monitorine instrumentation channels shovn in Table 3.21-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.22.1.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance-with the 0FFSITE DOSE CALCULATION MANUAL (ODCH).

APPLICABlull: At all times

• ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluent monitored by the affected channel or declare the channel inoperable,

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels SPERABLE, take the ACTION shown in Table 3.21-1. Exert best efferts to return the instrumentation to OPEFABLE status within 30 days and, if unsuccessful, explain in the next Annual Effluent Release Report why the inoperability was l not corrected in a timely mannar.

• For FT-84, and RM-L6, operabil.ty is not required when discharges are positively controlled through the closure of WDL-V257.

• For RN-L12 and associated IWTS/lWFS flow interlocks, operability is not required when discharges are positively controlled through the closure of IW-V72, 75 and IW-V280, 201.

• For FT-146, operability is not required when discharges are positivaly contro'lled through the closure of WDL-V257, IW-V72, 75 and IW-V280, 281.

BASES The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquia effluent during actual or potential releases. The alarm / trip setpoints far these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the lieits of 10 CFR Part 20. ,

l 3-96 Amendment No. /E, SS, 137

- 3.21.2 -EAD10 ACTIVE _GBSE00S PROCESS AND EFFLUENT MONITORING INSTRUMENTAL. LDH LIhlTING CONDITION FOR OPERATION 3.21.2 The radioactive gaseous process and effluent monitoring instrumentation chaanel: slown in Table 3.21-2 shall be OPERABLE with their alarm / trip s?tpoints set to ensure that the limits of Specification 3.22.2.1 are not exceeded. The alarm / trip sotpoints of these channels shall be determined in accordance with the 0FFSITE DOSE CALCULATION MANUAL (0DCH).

APPLICABILITY: As shown in Table 3.21-2.

ACTION:

a. With a radioactive gaseous process or effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification. immediately suspend the release of radioactive effluent monitored by the affected channel or declare the chamiel inoperable,

b. With less than the minimum number of radioactive gaseous process or effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.21-2. Exert best efforts to return the instrumentation to OPERABLE status within 30 days and, if unsuccestful, explain in the next Annual Effluent Release Report l why the inoperability was not correrted in a timely manner.

BASES The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluent during actual or potential releases. The alarm / trip setpcints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.

The inw range condenser offgas noble gas activity monitors also- provide data for determination of steam generator primary to secor 'ary leakage rate.

Channel operability requirements are based on an ASLB Drder No. LBP-84-47 dated October 31, 1984, and as cited in 20 NRC 1405 (1984).

3-100 Amendment No. 72, Jp), J)7, 157 .

c. To be representative' of the quantities and concentrations of radioactive

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materials in _ liquid effluent, samplos shall be collected continuou' sly in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be thorou-bly mixed in order for the composite sample to be representative of M *1 fluent release,

d. A batch release is the discharge of liquid wastes cf a discrete volume.

Prior to sampling for analyses, each batch shall be isolated, and be thoroughly mixed, by a~ method described in the ODCM,'to assure l representative sampling

e. A continuous release is the discharge of liquid wastes of a non-discrete volume; e.g., from a volume or system that' has an input flow during the continuous release.

f. The principal _ gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54,-Fe-59, Co-58, C0-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. TMs list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiabia, together with those of the above nuclices, shall also be analyzed and reported in the Annual Radioactive Effluent- l Release Report pursuant to TS 6.9.4.

I 4-99 Amendment No. 77, 137 l

Table 4.22-2 (Continuedl TABLE NOTATION

d. Chai :oal cartridges and particulate filters shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing (or after removal from sampler),

e. Tritium grab sample; shall be taken weekly from the spent fuel pool area whenever spent fuel is in the spent fuel pool.

f. The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications 3.22.2.1, 3.22.2.2, and 3.22.2.3.

g. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135 and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58,_Co-60, Zn-65, Mo-99, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be.

considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent Release Report pursuant to TS 6.9.4. -l

h. Applicable only when condenser vacuum is established. Sampling and analysis shall also be performed following shutdown, startup, or a '

THERMAL POWER change exceeding 15 percent of. RATED THERMAL POWER within one b5,r unless (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3.

1. Gross Alpha, Sr-89, and Sr-90 analyses do not apply to the Fuel Handling Building ESF Air Treatment System.

l j. If the Condenser Vent Stack Continuous lodine Sampler is unavailable, then alternate sampling equipment will be plat ed in service within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

l k. Applicable only when condenser vacuum is established.

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i 4-105 Amendment No. U , M , U p, 137

6.5.4 ANNUAL RADI0 ACTIVE EFFLVENT RELEASE REPORT 6.9.4.1 The Annual Radioactive Effluent Release Report cofering the-operations of the unit during the previous 12 months of operation ,

shall be submitted within 60 days after January 1 of each year.

The Report shall include a summary of the quantities of radioactive liquid and gaseous ef fluent and solid i.aste released from the unit. The material provided shall be: (1) consistent with the objectives outlined in the ODCH and PCP; and, (2) in conformance with 10 CFR 50.36(a) and Section IV.8.1 of Appendix I to 10 CFR Part 50.

Note: A single submittal may be made for the station. The submittal should combine those sections that are common to both units at-the station.

I 6-18 Amendment No. 77, 77, 179, 137

6.13- PROCESS CONTROL PROGRAM (PCP) 6.13.1 GPU huclear Corporation initiated changes to the PCP:

1, Sha'.1 be submitted to the NRC in the Annual Radioactive Effluent R91 ease Report for the period in which the changes were made. -This submittal shall contain:

a. sufficiently detailed information to justify the changes without benefit of additional or supplemental information;

b. a determination that the changes did not reduce the overall cueifonunce of the solidified waste product to existing criteria for solid wastes; and,

c. documentation that the changes have been reviewed ..id approved pursuant to 6.8.2.

2. Shall become effective upon review and approval by GPUNC Management, i

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k 6-23 Amendment Nos. JJ, ;iE, 77, 77, Jp6, Jp/,129 b

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6.14 0FFSITE-DOSE CALCULATION MANUAL (ODCM) 6.14.1 The ODCM shall be . . foved by the Comission prior to implementation.

6.14.2 GPU Nuclear Corporation initiated changes to the ODCH:

1. Shall be submitted to the NRC in the Annual Radioactive Effluant Release Report for the period in which the changes were made. This submittal s!.all contain:

a. sufficiently detailed information to justify the changes without benefit of additional or supplemental information;

b. a determination that the changes did not reduce the accuracy or reliability of dose calculations or setpoint determinations; and

c. documentation that the changes have been reviewed and approved pursuant to 6.8.2.

2. Shall become effective upon review and approval by GPUNC Management.

6.15 DELETED 6.16 POST-ACCIDENT SAMPLING PROGRAMS NUREG 0737 (II.B.3. II.F.1.2)

Program which will ensure the capability to accurately _ sample and analyze vital areas under accident conditions have been implemented.

The following programs have been established:

1. Iodine and Particulate Sampling

2. Reactor Coolant System -*

3. Containment Atmosphere Sampling Each program shall be maintained and shall include 'the following:

1. Training of personnel, 2, Procedures, and

3. Provisions for maintenance of sampling ano analysis equipment. >

6-24 Amendment No. 77, 77, 102 t

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