Text

Amend 108 to License DPR-50,revising Limits on Primary Coolant Activity & Incorporating Addl Restrictions on Vent/Purge Valve Operability & Surveillance
	ML20127D809
Person / Time
Site:	Crane
Issue date:	05/08/1985
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Metropolitan Edison Co, Jersey Central Power & Light Co, Pennsylvania Electric Co, GPU Nuclear Corp
Shared Package
ML20127D814	List: ML20127D809; ML20127D830;
References
	DPR-50-A-108 NUDOCS 8505170046
	Download: ML20127D809 (27)
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k UNITED STATES NUCLEAR REGULATORY COMMISSION

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METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY GPU NUCLEAR CORPORATION DOCKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATION, UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.108 License No. DPR-50 1.

The Nuclear Regulatory Commission (the Comission) has found that:

A. The application for amendment by GPU Nuclear Corporation, et al (the licensees) dated November 24, 1983, as revised and supplemented June 5, 1984, and December 3, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public,.and (ii) that such activities will be conducted in compliance with the Comission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security.or to the health and safety of the'public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

8505170046 850508 PDR ADOCK 05000289 P

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2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.c.(2) of Facility Operating License No. DPR-50 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.108, are hereby incorporated in the license. GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance, and revised procedures are to be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION J n F. Stolz, Chi f erating Reactors Branch #4 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: May 8, 1985 D

ATTACHMENT TO LICENSE AMENDMENT NO.108 FACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment

. number and contain vertical lines indicating the area of change.

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faggs iii 3-8 3-9 3-9a(Newpage) 3-9b (New page) 3-10*

3-41 3-41a 3-41b (New page) 3-41c (New page) 3-62a 3-62b 4-Sa 4-9 4-10 4-34 4-34a 4-34b 4-37 4-38 4-55b 4-55c 4-76 6-17

0verleaf page included for document completeness.

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TABLE OF CONTENTS Section

' Pace 3.16 SHOCK SLPRE5 SORS (SNUBBERS) 3-63 3.17 REACTOR BUILDING AIR TEbPERATURE 3-80 3.18 FIRE PROTECTION 3-86 3.18.1 FIRE DETECTION INSTRUMENTATION 3-86 3.18.2 FIRE SLPPRESSION WATER SYSTEM 3-88 3.18.3 DELUGE /SPRItKLER SYSTEMS 3-89 3.18.4 CO2 System 3-90 3.19 CONTAIPNENT SYSTEMS 3-95 3.20 SPECIAL. TEST EXCEPTIONS 3-95a 3.20.1 LOW POWER NATURAL CIRCULATION TEST 3-95a 3.21 RADI0 ACTIVE ENVIRONMENTAL SPECIFICATIONS 3-96 3.21.1 RADIDACTIVE LIQUID t.rrLUENT INSTRUMENTATION 3-96 3.21.2 RADI0 ACTIVE GASEOUS PROCESS APO EFFLUENT 3-100 MONITORING INSTRUDENTATION 3.22.1.1 LIQUID EFFLUENTS 3-106 3.22.1.2 OOSE 3-107 3.22.1.3 LIQUID WASTE TREATMENT 3-109 3.22.1.4 LIQUID HOLDUP TAPES

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3-110 3.22.2.1 00SE RATE 3-111 3.22.2.2 00SE, NOBLE GAS 3-112 3.22.2.3 DOSE, RADIDIODINES, RADI0 ACTIVE MATERIAL IN 3-113 PARTICULATE FORM APO RADIONUCLIDES OTHER THAN NOBLE GASES 3.22.2.4 GASEDUS RA0 WASTE TREATMENT 3-115 3.22.2.5

. EXPLOSIVE GAS MIXTURE 3-116 3.22.2.6 GAS STORAGE TANKS 3-117 3.22.3.1 SOLID RADI0 ACTIVE WASTE l

3.22.4 TOTAL DOSE 3-118 3-119 3.23.1 MONITORING PROGRAM 3-120 3.23.2 LAPO USE CENSUS 3-125 3.23.3 INTERLABORATORY C0hPARISON PROGRAM 3-127

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4 SURVEILLANCE STANDARDS 4-1 4.1 OPERATIONAL SAFETY REVIEW 4-1 4.2 REACTOR COOLANT SYSTEM INSERVICE INSPECTION 4-11 4.3 TESTING FOLLOWING CPENING OF SYSTEN 4-28 4.4 REACTOR BUILDING 4-29 4.4.1 CONTAINMENT LEAKAGE TESTS 4-29 4.4.2 STRUCTURAL INTEGRITY 4-35 4.4.3 DELETED i

4-37 1 t

4.5 EMERGENCY LOADING SEQUENCE AND POWER TRANSFER, EMERGENCY l

4-39 CORE C0 CLING SYSTEM Ato REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4.5.1 EMERGENCY LOADING SEQUENCE 4-39 4.5.2 E>ERGENCY CORE COOLING SYSTEM 4-41 4.5.3 REACTOR BUILDING COOLING AND ISOLATION SYSTEM 4-43 4.5.4 DECAY HEAT REMOVAL SYSTEM LEAKAGE 4-45 4.6 EFERGENCY PCWER SYSTEM PERIODIC TESTS 4-46 i

111 l-Amendment Nos. 77,$1,108

3.1.4 REACTOR COOLANT SYSTEM ACTIVITY J

3.1.4.1 LIMITING CONDITION FOR OPERATION The specific activity of the primary coolant shall be limited to; Less than or equal to 1.0 microcurie / gram DOSE EQUI; VALENT I-131, a.

and b.

Less than or equal to 100/T microcuries/ gram.*

. 3.1.4.2 APPLICABILITY: at all times except refueling.

3.1.4.3 ACTION:

MODES: Power Operation. Start-up, Hot Standby a.

With the specific activity of the primary coolant greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of'the line) shown on Figure 3.1-2a, operation may continue for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ** provided that the cumulative operating time under these circumstances does not exceed 800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months 'in any consecutive 12 month period during any fuel cycle.

With the total cumulative operating time at a primary coolant specific activity greater than 1.0 microcurie / gram DOSE EQUIVALENT 3

i I-131 exceeding 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months in any consecutive 6 month period during any fuel cycle, prepare and submit a Special Report to the L

Commission pursuant to Specification 6.9.3 within 30 days indicating the number'of hours of operation above this limit.

b.

With the specific activity of the primary coolant greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months **

during one continuous time interval or exceeding the limit line shown on Figure 3.1-2a, be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Power operation may continue when DOSE EQUIVALENT I-131 is below 1.0 microcuries/ gram.

With the specific activity of the primary coolant greater than 100/E c.

4 microcuries/ gram be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Power opera _ tion may continue when primary coolant activity is less than 100/E microcuries/ gram.

MODES: at all times except refueling.

d.

With the specific activity of the primary coolant greater than 1.0 i

microcurie / gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries/ gram perform the sampling and analysis requirements of Table 4.1-3 until the specific activity of the primary coolant is restored to within its limits. A Report shall be prepared and r

submitted to the Comission within 30 days. This report shall contain the results of the' specific activity analyses together with the following information:

I shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the dverage beta and gamma energies per disintegration (in MeV) for isotopes, other than todines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

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- The time period begins from the time the sample is taken.

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- Amendment No.108 l

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Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded, 2.

Fuel burnup by core region, 3.

Clean-up flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded, 4.

History of de-gassing operations, if any, starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded, and 5.

The time duration when the specific activity of the primary

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coolant exceeded 1.0 microcurie /gran DOSE EQUIVALENT I-131.

BASES The. limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will be well

within the Part 100 limit following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative, in that the specific site parameters of TMI-1, such as site boundary, location and meteorological conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.1-2a, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

Operation with specific activity levels exceeding 1.0 microcurie / gram DOSE EQUIVALENT I-131 but within the' limits shown on Figure 3.1.2a must I

be restricted t'o no more than 800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months per year (approximately 10 percent of'the units yearly operating time) since the activity. levels allowed by Figure 3.1-2a increase the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose at the site boundary by a factor of up to 20 following a postulated steam generator tube rupture. Reporting any cumulative operating time over 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months in any 6 consecutive month period with greater than.1.0 microcurie / gram DOSE EQUIVALENT 'I-131 will alert the NRC to the situation and allow.

sufficient time for evaluation and appropriate action before reaching the-800 hour limit.

Proceeding to HOT SHUTDOWN prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3-9 Amendment.No. 108

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The NRC staff has performed a generic analysis of airborne raciation released via the Reactor Building Purge Isolation Valves. The dose contribution due to

- the radiation contained in the air and steam releasec through the purge isolation valves prior to closure was found to be acceptable'proviced that the requirements of Specifications 3.1.4.1, 3.1.4.2 and 3.1.4.3 are ~ met.

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3.1. 5.

. CHEMISTRY

- Applicability Applies to acceptable concentrations of impurities for continuous operation Lef the reactor.

Objective-To protect the reactor coolant system from the effects of impurities.

Soeci fi cation 3.1.5.1 If the concentration of oxygen in the primary coolant exceeds 0.1 ppm during power operation, corrective action shall be initiated within eight hours to return oxygen levels to 10.1 ppm.

3.1. 5.2

. If _ the concentration of chloride in the primary coolant exceeds 0.15 ppm during power operat' ion, corrective action shall be initiated within eight hours to return chloride levels to 10.15 ppm.

,- 3.1. 5. 3~

-If the concentration of fluorides in the primary coolant exceeds 0.10 ppm following modifications or repair to the primary system involving velding, corrective action shall be initiated within eight hours to return fluoride levels to 10.10 ppm.

3.1.5.h If the concentration limits for oxygen, chloride or fluoride given in 3.1.5.1, 3.1 5.2, and 3.F.5.3 above are not restored within 2h hours of detection, the reactor shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter. If the normal operational limits.are not restored within an additional 2k-hour period, the reactor shall be placed in a cold shutdown condition within 2h hours thereafter.

3.1.5.5-If the oxygen, chloride, or fluoride concentration of the primary coolant system exceeds 1.0 ppm the reactor shall be brought to the hot shutdown condition using normal shutdown procedure and action is to be taken to return the system to within normal operation specifications.

If normal operating specifications have not been reached in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , the reactor vill then be brought to a cold shutdown condition.

B ses

.By maintaining the chloride, fluoride, and oxygen concentration in the reactor coolant within the specifications, the integrity of the reactor coolant system is protected against potential stress corrosion attack.(1,2)

Th2 oxygen concentration in the reactor coolant system is normally expected to be below detectable limits since dissolved hydrogen is used when the reactor is critical. The requirement that the oxygen concentration not exceed 0.1 ppm during pov { operation is added assurance that stress corrosion cracks will not occur.

If the oxygen, chloride, or fluoride limits are exceeded, measures can be taken to correct the condition (e.g., switch to the spare demineralizer, replace the ion ' exchange resin, or increase the hydrogen concentration in the makeup tank).

3-10

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3.6 REACTOR BUILDING' Applicability Applies to the containment integrity of the reactor building as specified below.

Objective To assure containment integrity.

Specification 2

3.6.1 Containment integrity as defined in Section 1.7, shall be maintained whenever all three of the following conditions exist:

a..

Reactor coolant pressure is 300 psig or greater.

b.

Reactor coolant temperature is 2000 F or greater.

c.

Nuclear fuel is in the core.

3.6.2 Containment integrity shall be maintained when both the reactor

~ coola'nt system is open to the containment atmosphere and a shutdown margin exists that is less than that for a refueling shutdown.

3.6.3 Positive reactivity insertions which would result in a reduction in shutdown margin to less than 1% A k/k shall not be made by control rod motion or boron dilution unless containment integrity is being maintained.

3.6.4 The reactor shall not be critical when the reactor building internal j

pressure exceeds 2.0 psig or 1.0 psi vacuum.

3.6.5 Prior to criticality following refueling shutdown, a check shall be made to confirm that all manual containment isolation valves which should be closed are closed and are conspicuously marked.

3.6.6

' While the reactor is critical, if a reactor building isolation valve (other than a purge' valve) is determined to be inoperable in a position other than the required position, the other reactor building isolation valve in the line shall be tested to insure OPERABILITY.

l If the inoperable valve is not restored within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the OPERABLE valve will be closed or the reactor shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to the COLD SHUTDOWN condition within an additional 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.6.7 The hydrogen recombiner shall be operable during REACTOR CRITICAL, HOT STANDBY and POWER OPERATION. With the hydrogen recombiner inoperable, restore the recombiner to operable status or bring the reactor to HOT SHUTDOWN within seven (7) days.

3.6.8 While containment integrity is required (See TS 3.6.1), if a 48" reactor building purge valve is found to be inoperable perform either 3.6.8.1 or 3.6.8.2 below.

Amendments Nos. 57, Jp/,108 3-41

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3.6.8.1 If inoperability is due to reasons other than excassive combined leakage close the associated valve and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months verify that the associated valve is OPERABLE. Maintain the associated valve closed until the faulty valve can be declared operable. If neither purge valve in the penetration can be declared OPERABLE within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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3.6.8.2 If inoperability is due to excessive combined leakage _ (See T.S.

4.4.1.7.1), within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months restore the leaking valve to OPERABILITY or-perform either a or b below.

a.

Manually close both associated reactor building isolation valves and meet the leaka either (1) or (2) ge criteria of T.S. 4.4.1.7.1 and perform below.

(1) restore the leaking valve to OPERABILITY within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

(2) maintain both valves closed by administrative controls.

verify both valves are closed at least once per 31 days and perform the interspace pressurization test of T.S.

4.4.1.7.1 every 3 months.

In order to accomplish repairs one containment purge valve may be opened for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following successful completion of an interspace pressurization test.

F b.

Be in HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.6.9 Except as specified in 3.6.11 below, the Reactor Building purge isolation valves (AH-V-1A&D) shall be limited to less than 310 and (AH-V-1B&C) shall be limited to less than 33' open, by positive means, while purging is conducted.

3.6.10 During STARTUP, HOT STANDBY and POWER OPERATION:

a.

Containment purging shall not be performed for temperature or humidity control.

b.

Containment purging is permitted to reduce airborne activity in order to facilitate containment entry for the following reasons:

(1) Non-routine safety-related corrective maintenance.

(2) Non-routine safety-related surveillance.

(3) Performance of Technical Specification required surveillances.

(4) Radiation Surveys.

I (5) Engineering support of safety-related modifications for pre-outage planning.

Amendment No. $7,108 3-41a f

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3 (6) Purging prior to shutdown to prevent delaying of outage commencement (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to shutdown).

. Containment purging is permitted for Reactor Building pressure c.

control.

d.

To the extent practicable the above containment entries shall be scheduled to coincide, in order to minimize instances of purging.

'3.6.11

- When the reactor is in COLD SHUTDOWN.or REFUELING SHUTDOWN continuous purging is permitted with the Reactor Building purge isolation valves opened fully.

Bases The Reactor. Coolant System conditions'of cold shutdown assure that no steam will be formed and hence no pressure will build up in the containment if the

' Reactor Coolant. System ruptures.

The selected shutdown conditions are based on the type of activities that are C

being carried out and will preclude criticality in any occurrence.

A condition requiring integrity of containment exists whenever the reactor coolant system is open 'to the atmosphere and there is insufficient soluble poison in the reactor coolant to maintain the core one percent suberitical in the event all control rods are withdrawn.

1The. reactor building is designed for~ an internal pressure of 55 psig, and an external pressure 2.5 psi greater than the internal pressure.

Due to industry reports of elastomer degradation in containment purge valve -

seats unique action requirements are now designated to help preclude common-mode' failure of both valves in series. An increased frequency of-leak rate-testing is also incorporated to help assure timely discovery and resolution of any seat degradation.

'An analysis of theJimpact of purging on ECCS performance and an evaluation of the _ radiological consequences of a design basis accident while purging have been' completed and accepted by the NRC staff. The purge isolation valves have been analyzed capable of closing against the dynamic forces associated with a loss-of-coolant accident when limited to a nominal 30' open.

-Allowing purge operations during STARTUP, HOT STANDBY and POWER OPERATION (T.S. - 3.6.10) is more beneficial than requiring a cooldown to cold shutdown from the standpoint of (a) avoiding unnecessary thermal stress cycles on the reactor coolant system and its components and _(b) reducing the potential for causing unnecessary challenges to the reactor trip and safeguards systems.

The recombiner unit is capable of controlling the expected hydrogen generation associated with 1) zirconium-water reactions 2) radiolytic decomposition of water. and 3) corrosion of. metals within containment. The recombiner is designed.in accordance with the recommendations of Regulatory Guide 1.7,

" Control of Combustible Gas Concentrations in Containment Following a LOCA",

LMarch 1971, the acceptance criteria of the Standard Review Plan (S.R.P.)

6.2.5., and NUREG 0578, July' 1979. In addition to the installed hydrogen recombiner, a _ second recombiner including. all piping, electrical, and 1 structural provisions is available on site.

Amendment No.'$7,'108 3-41b

' The hydrogen mixing is provided by the reactor building ventilation system to ensure adequate mixing of the containment atmosphere following a LOCA. This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit.

Interspace pressurization leak testing of containment purge valvds is performed once very three months. The primary objective of this testing per NRC Safety Issue B-24, is to identify excessive degradation of the resilient seats in a timely manner. Upon failing the quarterly test, manual closure of the valve and retesting are performed in order to identify leakage caused by excessive seat degradation. Manual closure means closure of the valve by means other than the normal operator.

REFERENCES FSAR Section 5.2.2.4.3 Amendment No. 57,108 3-41c v

3.15.2 REACTOR BUILDING PURGE AIR TREATMENT SYSTEM Applicability Appliestothereactorbuildingpurgeairtreatmentsystemanditsassociated filters.

Objective To specify minimum availability and efficiency for the reactor building purge air treatment system and its associated filters.

Specification 3.15.2.1 Except as specified in Specification 3.15.2.3 below, the Reactor Building Purge Air Treatment System filter AH-F1 shall be operable as defined by the Specification below at all times when containment integrity is required unless the Reactor Building purge isolation valves are closed.

3.15.2.2 a.' The results of the in-place DCP and halogenated hydrocarbon tests at maximum available flows on HEPA filters and charcoal adsorber banks for AH-F1 shall show less than 0.05% DCP penetration and less than 0.05% halogenated hydrocarbon penetration, except that the DCP test will be conducted with prefilters installed.

b.* The results of laboratory carbon sample analysis for the reactor building purge system filter carbon shall show greater than or equal to 90% radioactive methyl iodide decontamination g

efficiency when tested at 2500F, 95% R.H.

l 3.15.2.3 From and after the date that the filter AH-F1 in the reactor building purge system is made or found to be inoperable as defined by Specification 3.15.2.2 above, the Reactor Building purge

~

isolation valves shall be closed until the filter is made operable.

Not required until criticality for Cycle 5 operation.

Bases l

The Reactor Building Purge Exhaust System filter AH-F1 is normally used to i

filter all reactor building exhaust air. It is necessary to demonstrate operability of the filters to assure readiness for service if required to i

mitigate a fuel handling accident in the Reactor Builcing and to assure that 10CFR50 Appendix I limits are met. Reactor Building purging is required to be terminated if the filter is not operable.

i Amendments Nos. 55, $7, 7p,108 l

3-62a t

l

High efficiency particulate absolute (HEPA) filters are installed before the charcoal absorbers to prevent clogging of the iodine adsorbers for all emergency air treatment systems. The charcoal adsorbers are installed to reduce the potential release of radiciodine to the environment. If the efficiericles of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the 10 CFR 100 guidelines for the accident analyzed in FSAR update Section 14.2.2.1 which assumes 90% efficiency for inorganic lodines and 70K efficiency for organic iodines.

The flow through AH-F1 can vary from 0 7M to 50,000 CFM, the maximum purge flow rate. During all modes except COLD SHUTDOWN, the purge valves are limited to no more than 308.open (900 being full open). This provides greater assurance of contairrnent isolation dependability per NUREG 0737 Item II.E.4.2

- Item (2)(a). Makeup air is provided between filter AH-F1 and fans AH-E7A and B.

(See also T.S. 3.6).

The in-place DOP and halogenated hydrocarbon tests of the filter banks and the laboratory tests of the carbon samples will be done using the test methods and acceptance criteria of Regulatory Guide 1.52 (Rev. 2), except that DCP and Freon tests.will be performed such that radiation release limitations are not exceeded.

References (1) FSAR Sectice 5.3.3 (2) FSAR Sectie. 5.6 i

(3) FSAR Sectic.; 9.8 (4). Update FSAR Section 14.2.2.1 I

l l

Amendment No. JJ,108 3-62b r

(

~

l

(,

1ABLE 4.1-1 (Continued)

E CllANNEL DESCRIPfl0N Ct1ECK TEST CAllBRATE REMARKS 3

S 28. Radiation Monitoring Systems

W(1)(3)

M(3)

Q(2)

(1) Using the installed check source when background is less than twice g

the expected increase in cpm which

- would result from the check source g-alone. Background readings greater than this value are sufficient in g

themselves to show that the monitor is functroning.

M-(2) Except area gamma radiation monitors RM-G5, RM-G6, RM-G7, and RM-G8 which g-are located in the Reactor Building. When purging is permitted per T.S. 3.6, RM-G5 will be calibrated quarterly.

If purging is c.

j 0.

not permitted per T.S. 3.6 RM-G5 i

shall be calibrated at the next scheduled reactor shutdown following -

the quarter in which calibration would normally be due.

RM-G6, RM-G7, and RM-G8 which are in high radiation areas shall be calibrated at the next scheduled reactor shutdown following the quarter in which calibration is due,' if a shutdown during the quarter does not i

occur.

(3) Surveillances are required to be performed only when containment integrity is required? This applies to monitors which initiate containment isolation only.

29. ifigh arwl Low Pressure i

Injection Systems:

F1ow ChanneIs NA NA R

Does not include the monitors covered under specification 3.5.5.2 and 4.1.3

TABLE 4.1-3 9,.

e Y

MINIMUM SAMPLING FREQUENCY t

$r+

Item Check Frequency in 1.

Reactor Coolant a.

Specific Activity Determin-At least once each 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during -

ation to compare to the POWER OPERATION. HOT STANDBY.. START-100[EpCi/gm limit wo IIP, and HOT SHUTDOWN.

b.

Isotopic Analysis for DOSE 1)

I per 14 days during power-g EQUIVALENT I-131 Concentra.

operations.

tion

11) One Sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a one hour period during power operation, start-up and hot standby.

iii)

Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever the specific activity exceeds 1.0 pCi/

gram DOSE EQUIVALENT I-131 or 100/Epci/ gram during all modes but refueling.

p c.

Radiochemical for E 1 per 6 months

during power e

Dete rmination operation d.

Chemistry (C1, F and 0 )

2 5 times / week when T,yg is greater than 200'F.

e.

Boron concentration 2 times / week f.

Tritium Radioactivity Monthly 2.

Borated Water Storage Boron concentration Weekly and after each makeup when Tank Water Sample reactor coolant system pressure is greater than 300 is greater than 200'psig or T,yg F.

3.

Core Flooding Tank Boron concentration Monthly and after each makeup when RCS Water Sample pressure is greater than 700 psig.

e e

~

.t (i

Item Check i

Frequency 4..

Spent Tuel Pool Boron concentration Monthly and af ter each makeup h

Water Sample m-5.

Secondary Coolant a.

Gross activity

. Weekly when reactor coolant

=

system pressure is greater than i

8 300 psig or Tav ~ is. greater ithan 2000F.

w b.

Iodine Analysis **

5 6.

Boric Acid Mix Tank Boron concentration Twice weekly ***

3 or Reclaimed Boric Acid Tank o

7,8,9.

Deleted

10. Sodium Hydroxide Tank Concentration Quarterly and after each makeup.

11. Deleted

12. Condenser Partition 1 131 Partition Factor Once if primary / secondary Factor leakage develops, i.e., Gross e

E Beta-Gamma on secondary side of OTSG is greater than 2 x 10-8 microcuries per cc and evidence of fission products is present.

Until the specific activity of the primary coolant system is restored within its limits.

Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since the reactor was last subscritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

9 When the gross activity increases by a factor of two above background, an iodine analysis will be made and performed thereafter when the gross activity increases by 10 percent.

The surveillance of either the Boric Acid Mix Tank or the Reclaimed Boric Acid Tank is not necessary when that respective tank is empty.

4.4.1.2.4 Corrective A: tion ano Retest If at any time it is ceterminec that the criterion of 4.4.1.2.3 a.

above is exceeded, repairs shall oe initiated funediately, b.

If conformance to the criterion cf 4.4.1.2.3 is not cemonstrated within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , following date: tion of excessive local leakage, the reactor shall be shutdown an: depressurized until recairs

+

are effected ano the local leakage meets tne acceptance 4

criterion as oemonstrated by retest.

L.

t4.4.1.2.5 Test Frequency.

Local leak detection tests shall be perfer9ed at a frecuency of at least 'each refueling period, except tnat:

a.

The equipment hatch and fuel transfer tube seals snall be tested every other refueling period but in no case at intervals greater than 3 years.

If they are opened they will be tested after being closed.

b.

The entire personnel and emergency airlocks shall be tested once every six months.

When the airlocks are opened during the interim between six monthftests, the airlock door resilient seals shall be tested within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the first of each of a series of openings. This requirement exists whenever containment integrity is reesired.

c.

The reactor building purge isolation valves shall be leak tested each refueling interval per 10 CFR 50 Appendix J, Item III.D.2.

d.

An interspace pressurization test (See T.S. 4.4.1.7.1) shall be performed for reactor building purge isolation valves every 3 months. This requirement is not in effect durfnq cold shutdown.

Readings of the rotameters in ea n manifolc of the cenetration e.

pressurization system shall De re::rce: at periodic intervals not to exceed three montns.

4.4.1.3 Isolation Valve Functional Tests Every tnree months, remotely operated reactor building isolation valves snall be stroked to the position required to fulfill their safety function unless such operation is not practical during plant operation.

The' valves not strokec every tnree months shall be stroked curing each refueling perict.

4.4.1.4 Annual Inspection A visual examination of the accessible interior and exterior surfaces of the containment structure and its components shall be performeo annually ano pric:

to any integrated leak test to uncover any evicence of cetericr.ation which may affect either the containment's structural integrity or leak-tightness.

Tne discovery of any significant ceterioration snall oe accomoanied by ccrrective a:tions in accord with acceptable procedures, noncestructive tests, and I-inspections, and local testing where practical, ::10: to the conou t of any integrated leak test. Such repairs shall be rep :tec as part of the test results.

}

4-34 Amendment No. D,108

t-a 4.4.'1.5 Reacter Suildino Modifications.

Any major modificatiod or replacement of cceponents affecting the reactor building integrity shall be followed by either an integratec leak rate test or t' local leak-test, as appropriate, and shall meet the acceptance criteria of 4.4.1.1.5 anc' E

4.4.1.2.3, respectively..

4.4.1.6 Operability of Access Hatch interlocks l'

At least once per 6 =v*hs the operability of the personnel and emergency

~ hatch door interlocks and the associated control rcom annunciator circuits' shall be de+= = W. If the interlock pemits both doors to be open.at the sans time or does not provide accurate status indication in the control i

rocm the interlock shall be declared inoperable.

'2.

During periods when contaAnment integrity is reovired and an interlock is inoperable, each entry and exit via that airlock shall be locally superviseo by, a member of the unit-operating maintenance or tecnnical staffs, to assure N

~

that only one door is open at any time and that both ocors are prc?erly closed-following use. A record of supervision anc verification of closure snall be maintained during periods of interlock inoperability in an appropriate station log.

3..

If an intericek is inoperable for more than 14 cays following determination of

_ inoperability, use of the airlock, except for emergency purposes, shall be suspended until-the interlock is returned to operable status.

4.'4.I.7

~

0cerability of Pu oe Valves 1.

A periodic pressurization of the purge valve inters; aces to 50.6 psig per T.S.

4.4.1.2.5d shall be performed to help assure timely detection and resolution of valve and/or actuator degradation. Tne accep;ance criteria is that total local leakage when updated for the new purge valve leakage shall be less. than

0. 6L.

See Tech Spec 3.6.8 for further action.

A 2.

Tne rubber seats on purge valves shall te visually examined eacn refueling interval to detect oegradation (e.g. crackin;, trittleness, etc.) ano to i

assure timely cleahing, lubrication, and seat reclacement. As a minim.n seats shall be replaced at the first refueling following 5 years of seat service.

Bases (l)

.Tne reacto-builoing is desdened fN: an internal cressure of.55 psig anc a steam-air mixture temperatu e of 2817. Pric: to initial operation, the centainmen; was' strength tested at 115 percent of design pressu e ano leak fate tested at the cesign pressure.

The containment was also leak tested prior to initial c?sration at approximately 50 percent of the oesign pressu e.

inese tests establisheo tne acceptance criteria of 4.4.1.1.3.

The performance of periodic integrated and 1ccal leakape ste tests curing tne plant lj fe provides a current assessment of poten:ial leakage from the containment

!in case of an accident that would pressurize the interior of the containment.

In Amendment-Nos. 77, 75, JJ,108.

4-34a

order to provide a realistic appraisal of the integrity of the containment unoer accident conditions "as found" Itcal 15akage rgsults must be documentec for corr:ction of th2 integratId leakage rate test results. Containmsnt isolation valves are to be closed in the normal manner prior to local or integrated leakage rate tests.

The minimum test pressure of 27.5 psig for the periccic integrated leakage rate 4

test is sufficiently high to provide an accurate measurement of,the leakage rate and it duplicates the pre-operational leakage rate test at the Teduced pressure.

The specification provides a relationship for relating the measureo leakage of air at the reduced pressure to the potential leakage of 55 psig. The minimun of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months was specified for the integrated leakage rate test to help stabilize conditions and thus improve accuracy and to better evaluate data scatter. Tne frequency of the periodic integrated leakage rate test is keyec to the refueling schedule for the reactor, because these tests can best be performed during refueling shutdowns.

The specified frequency of periodic integra:ed leakage rate tests is based on three major considerations. Firs: is the low probabill:7 of leaks in the 11 set, because of conformance of the complace conta h en: to a 0.10 percen:

1eakage rate a: 55 psig-during pre-operational :esting and the absence of any significan: s:resses in the liner during reac:or operation.

Second is the

-more' frequent res:ing, at design pressure, of those portions of the containment envelo.pe tha: are mos: likely to develop leaks during reactor operation, and the lov value (0.06 percen:) of leakage : hat is specified as acceptable fren penetra: ions and isclation valves. *hird is the tendon stress surveillance progran which p cvides assurance tha: an i=por an: par: of the s::ue: ural in:egri:7 of the con:ainmen: is nain: tined.

More frecuen: :es:ing cf various pene: a:icts is specified as these *.oca:1:ns are nere susceptible to leakage than the reae::: building liner due :c :he techanical closure involved. The basis for specifying a :etal leakage rate of 0.06 percen:

~fron :hese-pene::sticus and isola:ics valves is tha: more chan one-half of :he allevable in:egrated leakage ra:e vill be fron these sources.

Valve operability tests are specified to assure proper closure or opening of tne

~-

reactor building isolation valves to provide for isolation or functioning of l

Engineered Safety ~ Features systems. Valves will be strokea to the position required to fulfill their safety function unless it is established that such testing is not practical during operation. Yalves that cannot be full-stroke tested will be part-stroke tested during operation and full-stroke tested curing each normal refueling shutdown.

Periodic surveillance of the airlfck interlock system is specified to assure j

continued operability and preclude instances where one or both ocors are l

inaavertently left open. When an airlock is inoperable anc containnent integrity is required, local supervision of airlock operation is specified.

Purge valve interspace pressurization test operability requirements ano inspections provide a high cegree of assurance of purge valve performance as containment isolation barriers.

l References (1) FSAR, Section 5 4-34b Amendment Nos. #, 97,108 l

L

4, a

4.4.3 DELETED

~

l 4-37 i

Amendment No.108

e THIS PAGE INTENTIONALLY LEri E.ANK.

4-38 Amendment No. 108

4.12.2 REACTOR BUILDING PURGE AIR TREATMENT SYSTEv

,'Applic5bility Applies to the reactor building purge air treatment system and associated components.

Objective-To verify that this system and associated components will be able to perform its design functions.

Specification 4.12.2.1 _At least once per refueling interval or once per 2 years, whichever comes first it shall be demonstrated that the pressure drop across the combined EPA filters and charcoal adsorber banks is less than 6 inches of water at system design flow rate (1105).

i 4.12.2.2 a.* The tests and sample analysis required by Specification 3.15.2.2, shall be performed initially, once per refueling interval or 2 years, whichever comes first, or within 30 days prior to the movement of irradiated fuel in containment and

. following significant painting, steam, fire, or chemical release in any ventilation zone comunicating with the system that could contaminate the EPA filters or charcoal adsorbers, b.* DOP testing shall be performed after each complete or partial replacement of a EPA filter bank or after any structural maintenance on the system housing which could affect EPA frame bypass leakage, c.* Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of a charcoal adsorber bank or after any structural maintenance on the system housing which could affect the charcoal adsorber bank bypass leakage.

d.' The DOP and,halogenated hydrocarbon testing shall be performed T

at the maximum available flow considering physical restrictions, i.e., purge valve position, and gaseous radioactive release criteria.

e.

Each refueling, AH-E7A&B shall be shown to operate within + 5000 cfm of design flow (50,000 cfm) with purge valves fully open.

4.12.2.3 An air distribution test shall be performe:: on ths TEPA filter bank initially and after any maintenance or testing that could affect the air distribution within the system. Tne air distribution across the EPA filter bank shall be uniform within +20%.

The test shall be performed at 50,000 cfm (310%) flow rate with purge valves fully l

open.

Surveillance to be performed prior to Cycle 5 criticality in 1 eu of once per refueling interval or once per 2 years.

i Bases Pressure drop across the combined EPA filters and charcoal adsorbers of less than 6 inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter.

Pressure drop should be determined at least once every refueling interval to show system performance capability.

' Amendment Nos. 5), $$ 108 4-55b

The frequency of tests and sample analysis are necessary to show that the HEPA

.S

' filters and charcoal adsorbers can perform as evaluated. Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant shall be performed in accordance with approved test procedures. The charcoal adsorber efficiency test procedures should allow for the removal of one adsorber tray, emptying of one bed from the tray, mixing the adsorbent thoroughly and obtaining at least two samples.

Each sample should be at least two inches in diameter and a length equal to the thickness of the bed.

If test results are unacceptable all adsorbent in the system should be replaced with an adsorbent qualified according to Regulatory Guide 1.52, March 1978.

Tests of the HEPA filters with DOP aerosol shall also be performed in accordance with approved test procedures.

Any HEPA filters found defective should be replaced with filters qualified according to Regulatory Guide 1.52, March 1978.

Fans AH-E7A&B performance verification is necessary to ensure adequate flow to perform the filter surveillance of T.S. 4.12.2.1 and 4.12.2.3 and can only be demonstrated by running both fans simultaneously. This can only be accomplished when purge valves are not limited to 308 open (i.e., cold s hutdown).

Since H recombiners at TMI p purge has been superseded by the installation of H7

~

I..the reactor building purge exhaust system no longer is relied upon to serve an operating accident mitigating (i.e. LOCA) function. The retest requirement of T.S. 4.12.2.2a has therefore been changed to reflect the same retest requirements as the auxiliary and fuel handling building ventilation system which similarly serves no operating accident mitigating function.

4 If significant painting, steam, fire, or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated 'from the fumes, chemicals or foreign material, the same tests and sample analysis shall be performed as required for operational use. The determination of significant shall be made by the Operations and Maintenance Director - TMI-1.

4 l

4-55c l

Amendment Nos. 55, 108

..m

4.18.6 HOSE STATIONS Applicability:

Hose stations listed in Taole 3.18-2.

Objective:

To insure system operaDility.

Soecification:

4.18.6.1 Eacn fire hose' station shall be verifiec operable:

a.

At least once per month

by visual inspection of the station tc i assure all equipment is at the station.

b.

At least once per 18 months

by removing the hose for inspection l and re-rackirg, and replacing all gaskets in the couplings that are degraded.

c.

At least once per 3 years, partially open hose station valves to verify valve operability and no blockage.

d.

-At least once per 3 years by conoucting a hose hyorostatic test at a pressure at least 50 psi greater than the maximum pressure available at that hose station.

For hose stations in the Reactor Buildino these inspections may be deferred, if purging is not permitted per TS 3.6, until the first shutdown greater than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months following the interval which permits purging.

~.

~

4-76 Amendment No. M, 108

(2)

Steam Generator Tube Inspection witnin 3 months after Program (See Section 4.19.5) completion of insp;ction.

.. =

4 (3)

Containment Integrated witnin 6 months after Leak Rate Test completion of test.

(4)

Inservice Inspection Program within 6 months after five years of operation.

(5)

Radioactive Sealed Scurce Leakage Test within 90 days after revealing the presence of > 0.005 completion of test.

microcuries of Removable Contamination.

(6) k ecial Report - Exceeding 500 hrs. of submit within 30 days.

- operation with greater than 1.0 micro-curie / gram DOSE EQUIVALENT I-131 in any 6 month period. Indicate number of hours of operation above this limit.

See T.S. 3.1.4 6.9.4 ANNUAL RADIOLOGICAL ENVIROPe4 ENTAL OPERATING REPORT NOTE:

A single submittal may be made for the station.

The J: ~

submittal should combine those sections thet are common to both units at the station however, for units with separate radwaste systems, the submittal shall specify the release of radioactive material from each unit.

6.9.4.1 Routine radiological environmental operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to May.1 of each year.

6.9.4.2 The annual radiological environmental operating reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environnent. The reports shall also include the results of the land use censuses required by Technical Specification 3.23.2.

If harmful effects of evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem.

The annual radiological environmental operating reports shall include summarized and tabulated results in the format of the Radiological Assessment BTP on the REMP March 1978 of all radiological environmental samples taken during the report period.

In the event that some results are not available for inclusion with i

the report, the report shall be submitted noting and: explaining the reasons for the missing results.

The missing data sball be submitted as soon as possible in a supplementary report.

Tne reports shall also include the following:

a summary description of the radiological environmental monitoring program; a map of all sampling locations keyed to a table giving distances and directions from one reactor; and the results of licensee participation in the Amendment Nos. 57, $$, 72, 77,108 6-17 r

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