ML20155F908

Amend 62 to License DPR-45,revising Tech Specs Per Util 870930 Request as Revised on 880222

ML20155F908
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	05/31/1988
From:	Rubenstein L Office of Nuclear Reactor Regulation
To:
Shared Package
ML20155F899	List: ML20155F895 ML20155F908 ML20155F916
References
NUDOCS 8806170101
Download: ML20155F908 (40)

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d ~go UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION h a WASHINGTON, D. C. 20555 ,

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i DAIRYLAND POWER COOPERATIVE DOCKET NO. 50-409 LA CROSSE BOILING WATER REALTOR (LACBWR)

AMENDMENT TO PROVISIONAL LICENSE Amendment No. 62 License No. DPR-45

1. The Nuclear Regulatory Comission (the Comission) has found that:

A. The application for amendment filed by Dairyland Power Cooperative (the licensee), dated September 30, 1987 as revised February 22, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations as set forth in 10 CFR Chapter I; B. The facility will be maintained in conformity with the application, the provisions of the Act, and the regulations of the Comission; C. There is reasonable assurance: (1) 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 Commission's regulations set forth in 10 CFR Chapter I; 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 Comission's regulations and all applicable requirements have been satisfied.

8806170101 880531 DR ADOCK OSO g 9

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment, and Paragraph 2.C.(2) of Provisional License No. DPR-45 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 62, are hereby incorporated in the license.

The licensee shall maintain the facility in accordance with the Technical Spe:ifications.

3. This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULA*iORY COMMISSION 4

Lester'S, Rubenstein, Acting Director Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, Y and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

Changes to the Technical Specifications Date of Issuance: May 31, 1988 I

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ENCLOSURE TO LICENSE AMENDMENT NO. 62 PROVISIONAL LICENSE NO. DPR-45 DOCKET N0. 50-409 Replace the following pages and figures of the Appendix A Technical Specifications as indicated below. The revised pages and figures are identified by Amendment number and contain vertical lines indicating the area of change.

Remove Pages Insert Pages 27d 27d 27e 27e 27f 27f 27g 27g 27h 27h 271 271 271-27rr 271-27rr, ane page 28 28 29 29 29a 29a 29b 29b 30-32e 30-32e, one page 32f 32f 32r(l)-32r(5) 32r(1)-32r(5), one page 32t-33 32t-33, one page 34 34 35-36 35-36, one page 37 37 37n 37n 37o-37bb 37o-37bb, one page 43 43 44 44  ;

49 49 55 55 58 58 i 5-1 5-1 5-2 5-2 5-4 5-4 5-5 5-5 5-6 5-6 5-6a 5-6a 5-7 5-7 5 5-10 5 5-10, one page 5-11 5-11 5-12 5-12 l 5 5-18 5 5-18, one page 6-1 6-1 l

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4. OPERATING 1.TMITAT10NS 101 2 Definitions _ __

The following terms are defined so that uniform interpretation of these specifications may be achieved. When these terms appear in enpitalized type, t he following definit ions apply in these Technical Specifications.

ACTION ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.

CilANNEL CALIBRATION A CllANNEI. C6LIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNET CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CilANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by an series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

CilANNEL CHECK A CllANNEL CIIECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels l

measuring the same parameter.

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J 27d Amendment No. 62

4. OPERATING LIMITATIONS 4.0.1 Definitions - (cont'd)

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall bet

a. Analog channels - the injection of a simulated signal into the c.hannel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.

b. Bistable channels - the injection of a real or simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

CONTAINMENT INTEGRITY CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be isolated during accident conditions are either:

1. Capable of being closed by an OPERABLE containmeni automatic isolation valve system, or

2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position.

b. The freight door is closed,

c. Each air lock is OPERABLE,

d. The containment leakage rates are within the limit, and

e. The sealing mechanism associated with each penetration (e.g.,

welds, bellow, o-rings) is OPERABLE.

27e Amendment No. 62

4. OPERATING LIMITATIONS 4.0.1 Definitions - (cont'd) - - - .

DOSE EQUTVALENT I-13]

DOSE EQUIVALENT I-131 shall bo that concentration of I-131 uCi/ gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, 1-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculations of Distance Factors for Power and Test Reactor Sites."

E - AVERAGE DISINTEGRATION ENERGY 5 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 average beta and gamma energies per disintegration, in MeV, for isotopes other than iodines with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

EFFLUENT RELEASE BOUNDARY The Dairyland Power Cooperative property line within the 1109 ft. radius Exclusion Area is the EFFLUENT RELEASE BOUNDARY. See Figure 4/5.3.

FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table of Surveillance Frequency Notation. l GASEOUS RADWASTE SYSTEM A GASEOUS RADWASTE SYSTF,M is a system designed to reduce radioactive gaseous I effluents by collecting primary coolant offgases from the primary system and providing for delay, holdup or filtering for the purpose of reducing the total radioactivity prior to release to the environment.

1 27f Amendment No. 62

4. OPPHATING 1. IMITATIONS 1_. _0 ._1 Definitions (cnjf d) _. _ _ . . . . . _ . . . . . . . . . . . . .

ffCMBER(S) 0F THE PUBLIC MDIDER(S) 0F Tile PUBLIC shall include all persons who are not occupatinnally associated with the utility. This category does not include employees of the utility, its contractors or vendors. Also excluded from this categoryThis are persons who enter the site to service equipment or make deliveries.

CHiegory does include persons Who use portions of the site [or recreational or other purposes not associated with the utility. l OFFSITE_fjoSE CALCULATION MANUAL An OFFSITE DOSE CALCULATION MANUAL (ODCM) shall be a manual containing the methodology and parameters to be used for the calculation of offsite doses due to radioactive gnscous and liquid effluents and for the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints.

11. shall describe the radiological environmental monitoring program.

OPERABLE-OPERABILITY A system, subsystem, train, component or device shall be OPERADLE or have ,

OPERABILITY when it is capable of performing its specified function (s) and l l

when all necessary attendant instrumentation, controls, a normal and an emergency electrical power source, cooling or seal water, lubrication or  ;

other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing l their related support funct ion (s).

OPERATIONAL CONDITION - CONDITION An OPERATIONAL CONDITION, i.e. CONDITION, shall correspond to any one inclusive combination of power level and average reactor coolant temperature specified in Table of OPERATIONAL CONDITIONS.

27g Amendment No. 62

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4. OPERATING I, IMITATIONS 1

1.0.1 I)efini tions - (cont'd) _

PROCESS CONTR01. PRodilAM (PCP)

The PROCESS CONTR01. PitOGRAM shall contain the current formula, sampling, analyses, tests and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Part 20,10 CFil Part 71 and Federal and Stnte regulations and other rnquirements governing the disposal of the radioactive waste.

REPORTADLE EVENT A REPORTADI.E EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50. A Licensee Event Report shall be submitted for REPORTADl.E EVENTS.

RESTRICTED AREA A RESTRICTED AREA nhall be nny area within the exclusion boundary, access to which is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials. See Figure 4/5.3.

s 27h Amendment No. 62

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4. OPERATING LIMITATIONS 4.0.1 Definitions - (cont'd)

SOLIDIFICATION SOLID 1FICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.

SOflCE Cl!ECK A SothCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

STAGGERED TEST DASIS A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals.

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b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval. l UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area not controlled by the licensee for purposes of protection of MEMBERS OF THE PUBLIC from exposure to radiation and radioactive materials.

VENTILATION EXIIAUST TREATMENT SYSTEM A VANTILATION EXHAUST TREATMENT SYSTEM is any sys',em designed and installed to reduce radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through High E!lficiency Particulate filters for the purpose of removing p .rticulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents.

271 Amendment No. 62

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4.0.2 4.1.1 4.1.2 1

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(Pages 271 - 27rr) Amendment No. 62

- i 4.1 GENERAL 4.1.3 Deleted, f

4.1.4 Deleted.

I 4.1.5 Deleted. l 4.1.6 Deleted.

4.2 OPERATING LIMITS 4.2.1 Reactor Building .

4.2.1.1 CONTAINMENT INTEGRTTY shall be maintained during handling of irradiated fuel.

4.2.1.2 Gasketed closures and ventilatie>n system closures which have been subjected to maintenance, repair or other operations which might affect their performance shall, before any subsequent operation for which containment integrity is required, be tested for leak tightness using the soap-bubble technique (or other method of equivalent sensitivity). This test shall be performed using a pressure clifferential no less than 0.5 psi and the results shall be used as a guide in evaluating leakage.

28 Amendment No. 62.

'4.2.1.3 The number of containment vessel electrical penetrations may be varied, as may the containment vessel piping penetrations, provided the new penetrations are equivalent in design to the existing penstrations. New penetrations are defined as conduit or piping requiring attachment to the containment vessel shell. After any such penetration changes, a leak test meeting the requirements of Section 5.2.1.2 shall be performed , j 4.2.1.4 Existing containment vessel penetrations may be removed from or placed in service, provided the containment vessel is not affected and the closures are equivalent in strength and tightness to those previously installed. After any such change the penetration, exclusive of the cor.tainment shell connection, shall be tested for leak tightness using a test l meeting the requirements of Section 5.2.1.2.

4.2.1.5 The reactor building vacuum breakers shall be set to relieve at a differential (external-over-internal) pressure not exceeding 0.5 psi. ,

l 4.2.1.6 The reactor building ventilation system isolation dampers shall l close upon loss of control air or loss of electrical power supply, and they shall be automatically closed by abnormal conditions as specified in Table 1.

4.2.1.7 The main steam line, the reactor building vent header, and the decay  ;

heat system blowdown line shall be isolated automatically by abnormal conditions as specified in Table 1.

4.2.1.8 The containment building steel shell temperature shall be greater than 00F during reactor operation or during any integrated leak rate test l performed with a pressure exceeding 10.4 psig. l 4.2.1.9 Deleted. l l

29 Amendment flo. 62

REACTOR BUILDING CONTAINMENT VENTILAT1,0N DAMPE_RS LIMITING CONDITION FOR OPERATION 4.2.1.10 The containment ventilation inlet and outlet dampers shall be OPERABLE with isolation times of less than or equal to 10 seconds.

APPLICABILITY: Whenever CONTAINMENT TN1TGRITY (Specification 4.2.1.1) is required.

ACTION:

With one or more of the above ventilation damper (s) inoperable, )

A.

suspend fuel handling or isolate the affected penetration with an automatic valve secured in its closed position, or with a blind flange, within I hour.

B. The provisions of Specification 3.0.4 are not applicable if the affected penetration is isolated.

SITRVEILLANCE REQUIREMENTS 5.2.1.10.1 The ventilation dampers shall be demonstrated OPERABLE prior to ,

y returning the damper to service after maintenance, repair or replacement work l l

is performed on the damper or its associated actuator, control, or power circuit by performance of a cycling test, and verification of isolation time. l j

5.2.1.10.2 The isolation time of each above damper shall be determined to be l within its limit when tested quarterly. ,

5.2.1.10.3 The seat rings of the ventilation inlet and outlet dampers shall be replaced at least once per 5 years.

BASES The Nuclear Regulatory Commission requested that similar Technical Specifica-tions per Generic Item B-24 and NUREG 0737 Item II.E.4.2 be submitted to help assure operability of containment ventilation dampers. The ACTION statement has been modified due to the plant's permanent shutdown. DPC had previously committed to replace the containment ventilation inlet and outlet dampers' resilient sealing material at least once each 5 years until such time as additional in-situ data an be accumulated to justify a longer interval. If in-situ data is accumulated which supports a longer seal replacement inter-vals, a change to Specification 5.2.1.10.3 may be requested. Specification 3.0.4 is not applicable if the affected penetration is isolated since the safety function of the dampers is to close.

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29a Amendment No. 62

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I 4.2.2 Reactor Vessel. Coolant, and Auxiliary Systems .

l 4.2.2.1 Additional penetrations to the systems containing reactor coolant shall be designed, manufactured, and tested according to the provisions of the ASME Boiler and Pressure Vessel Code and the ASA Code for Pressure Piping applicable as of June 1962. These additional penetrations shall be limited to instrument connections and piping connections, the latter being no larger than 1-inch inside diameter.

4.2.2.2 The reactor coolant shall be light water and shall conform to the following requirements.

1 CONDITIONS 4 & 5 Normal Limit Chloride concentration . 5 ppm  :

pli 5.3 - 8.6 Conductivity 10 umho/cm The primary system chemistry para.,eters defined in this section shall bc  ;

determined at least once every 7 daye in Conditions 4 and 5. l 4.2.2.3 Deleted I

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. 29b Amendment No. 62

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4/5.2.2.21 4.2.2.22 5.2.16 l (Next page is Page 32e(l)) ,

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(Pages 30 - 32e) Amendment No. 62

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i (Page 32f) Amendment No. 62

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[Pages 32r(l) - 32r(S)] Amendment No. 62

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l (Pages 32t - 33) Amendment No. 62

4 4.2.5 Deleted l 4.2.6 Su_f_ety_ Instrument stiop 4.2.6.1 The safet y instrumentation shall provide isolation action and other l safety actions as specified in Table 1 of these specifications.

4.2.6.2 The setpoints for the safety instrumentation shall be as specified in Table 1. l 4.2.6.3 Key switches shall permit operational, maintenance, and test bypass of the safety instrumentation only with the approval of the Shift Supervisor. l 4.2.6.4 Deleted. l 4.2.6.5 Deleted. l 4.2.6.6 Deleted. l 4.2.6.7 Deleted. l 4.2.6.8 Safety channels directly backed up by an identical channel or channels may be bypassed for maintenance or testing. l 4.2.6.9 Deleted. l 4.2.7 Deleted. l 6

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34 Amendment No. 62

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l TIIIS PAGE INTENTIONALLY LEFT BLANK (Pages 35 - 36) Amendment No. 62

4.2.8 Spent Fuel Storage and Handling 4.2.8.1 Deleted.

4.2.8.2 Irradiated fuel elements shall be stored underwater in spent fuel storage racks that are positioned on the bottom of the spent fuel storage well, or in an approved shipping cask.

4.2.8.3 During the handling of irradiated fuel elements that have been operated at power levels greater than 1 Mwt the depth of water in the reactor upper cavity and/or the spent fuel storage well shall be at least 2 feet above the active fuel.

4.2.8.4 Deleted.

4.2.8.5 With the exception of a spent fuel shipping cask, the core spray bundle, the transfer canal shield plug and the other components and fixtures that are normally located and used within the spent fuel storage well, no objects heavier than a fuel assembly shall be handled over the spent fuel storage well.

37 Amendment tio. 62

PLANT SYSTEMS LIMITING CONDITIONS FOR OPERATION 4.2.23 At least one of the following Fuel Element Storage Well Water supplied should be OPERABLE:

a. The Demineralized Water Tank with a minimum water level of two feet, or

b. The Overhead Storage Tank with a minimum contained water volume of 5,000 gallons.

APPLICABILITY: At all times.

ACTION:

With neither the Demineralized Water Tank nor the Overhead Storage Tank OPERABLE, restore level in at least one tank within 7 days.

SURVEILLANCE REQUIREMENTS 5.2.23.1 The Demineralized Water Tank shall be demonstrated OPERABLE by verifying the minimum water level in the tank at least once per 7 days.

5.2.23.2 The Overhead Storage Tank shall be demonstrated OPERABLE by verifying the minimum contained water volume at least once per 7 days.

BASIS FOR FUEL ELEMENT S'iORACE WELL WATER SUPPLY

............ ___________________ . ....==;_ ____________

s A minimum Fuel Element Storage Well water supply has been established to ensure that sufficient water is available onsite to provide short-term makeup to the storage well while an alternate supply is being established if needed.

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37n Amendment No. 62

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l TABLE 4.3.2.1 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION ^

MINIMUM CilANNELS APPLICABLE OPERABLE CONDITIONS ACTION INSTHi! MENT

1. Reactor Containment Building Ventilation Monitor System

• B

n. Particulate Activity 1 Monitor

• B

b. Gaseous Activity Monitor 1

c. Sampler Flow Rate Measuring Device 1

• C

2. Stack Monitor System

a. Noble Gas Activity Monitor 1 ** D

b. Iodine Sanpler 1 ** E

c. Particulate Sampler 1 ** E

d. Sampler Flow Hate 1 ** C.

Measuring Device

3. Deleted. l

• When Containment Building Ventilation System is in operation.

• • At all times, unless alternate monitoring is available.

A. For post-accident instrumentation, refer to Section 4.5.2.

B. With the number of channels OPERABLE less than required by the Minimum Channels OPERABI.E requirement, effluent releases through this pathway may continue for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, as long as stack monitors are OPERABLE. l C. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may l continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

D. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided alternate monitoring is available or grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

E. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided alternate monitoring is available meeting the requirements of Table 5.3.2.1 or continuous collection of samples with auxiliary sampling equipment is initiated within I hour.

F. Deleted. l 43 Amendment No. 62

TABLE 5.3.2.1 RADIOACTIVE GASEOUS EFFLUENT MONITORI_NG,_INSTRINENTATION SURVEILLANCE REQUIREMENTS CilANNEL (4) SURVEILLANCE CHANNET, SOURCE FUNCTIONAL CHANNEL REQUIREMENT CHECK _ CRECK TEST CALIDRATION CONDITIONS _

INSTRUMENT

l. Reactor Containment Building Ventilation  ;

Monitor System

a. Particulate g

• Activity Monitor D M Ct2)

b. Gaseous Activity M Q(1) R *  ;

Monitor D

c. Sampler Flow Rate

• Measuring Device D N/A Q(3) R

2. Stack Monitor System <

a. Noble Gas Activity M Q(2) R

• f Monitor D

• l

b. Iodine Sampler D M Q(a) R Q(2) R *

c. Particulate Sampler D N/A

d. Sampler Flow Rate )

Q(3) R

• Measuring Device D N/A

3. Deleted. l

• During applicable conditions per Table 4.3.2.1.

l (1) The CilANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exist: f

a. Instrument indicates measured levels at or above the alarm setpoint. l

b. Instrument indicates a downscale failure (provides control room l

annunciation alarm only). l

c. Instrument indicates a circuit failure (provides control room i annunciation alarm only). . i l

(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditior.s exist:

a. Instrument indicates measured level above the alarm setpoint on one channel.

b. Instrument indicates a failure by n Low Flow and Low Count Rate signal.

c. Instrument controls in Maintenance mode.

(3) The CHANNEL FUNCTIONAL TEST shall also demonstrate that the control room local alarm occurs if the flow instrument indicates measured levels below the minimum and/or above the maximum alarm setpoint.

(4) The CHANNEL CALIBRATION shall be conducted in accordance with plant procedures.

44 Amendment No. 62

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49 Amendment No. 62

RADIOACTIVE EFFLUENTS BASES 4/5.3.2.4 p0SE. RADIONUCLIDES OTHEg_THAN NODLE GASES This specification is provided to implement the requirements of Sections II.C, III.A, IV.A and Annex of Appendix I, 10 CFR Part 50. The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III. A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and dnta such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated.

4/5.3.2.5 Deleted. l 4/5.3.2.6 VENTILATION EXHAUST TREATMENT SYSTEM (CONTAINMENT BUILDING)

The OPERABILITY of the ventilation exhaust treatment system (Containment Building) ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. This specification implements the requirements of 10 CFR Part 50, and the desien objectives given in Section II.D of Appendix I to 10 CFR Part 50.

4/5.3.3 SOLID RADIOACTIVE WASTE The OPERABILITY of the solid radweste system ensures that the system will be available for use whenever solid radwastes require processing and packaging prior to being shipped offsite. 1nis specification implements the requirements of 10 CFR Part 50.36a and General Design criterion 60 of Appendix A to 10 CFR Part 50.

4 /5,. 3. 4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR 190.

The specification requires the preparation and submittal of a Special Report l whenever the calculated doses from plant radioactive effluents exceed twice l the design objective doses of Appendix 1. The Special Report will describe a l l

course of action which should result in the limitation of dose to a real individual for 12 consecutive. months to within the 40 CFR 190 limits. l l

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i 58 Amendment No. 62 I

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. l S. MAINTENANCE l

5.1 GENERAL 5.1.1 Maintenance operations and routine tests shall be performed in l conformance with these specifications.

5.1.2 Maintenance operations shall be performed as authorir.ett by the Shif t Supervisor. Maintenance involving the opening of systems containing radioactive materials shall be conducted under the surveillance of a llealth Physics representative.

5.1.3 Deleted. l 5.1.4 Deleted.

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5-1 Amendment No. 62 2

5.1.5 Components which have been repaired, replaced, or otherwise subjected to temporary or permanent modification shall be tested in accordance with procedures which are appropriate in view of the nature of the repair, replacement or modification, and in view of the condition of the system.

5.1.6 Deleted.

5.1.7 Deleted.

5.2 TESTING 5.2.1 Containment Testing 5.2.1.1 Deleted.

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5-2 Amendment No. 62

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'. l the electrical penetrations

- the renctor building spray valve shaft penetration. l the freight door, the main steam line penetration, the feedwater line penetration, the heating steam line and condensate return penetration, the containment building airlocks, and ,

the flanges of the ventilation system inlet and exhaust ducts. i component Leak surveillance system: A leak survelliance system (i.e., .

continuous pressurization of individual containment components) that l maintains a pressure not less than 52 psig at individual teet chambers of l containment penetrations and seals during normal reactor operation shall be i acceptable in lieu of Type B tests for the components under such leak l surveillance. J (b) Type C Tests: Containment isolation valve leak detection testing shall be conducted at a pressure of 52 psig. 1 (c) Acceptanen criteria: The combined leakage rate for all  :

penetrations and valves subject to Type B and C tests shall be less than 60  !

percent of the design leak rate of 0.1 percent per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the Containment Building volume, i

(d) Corrective Actions: Leaks which cause the acceptance criteria of (c) to be exceeded shall be repaired and retested until the criteria is met, l Repairs of lesser leaks are optional. l 1

(e) Test Trequency: Type B tests (except for air locks and electrical 1 penetrations) shall be performed at intervals no greater than two years. Air locks shall be tested at 4-month intervals. The freight door shall be tested following each closure. Electrical penatrations shall be tested at intervals ng greater than one year. Type C tests shall be performed at inte-vals no dreater than two years.

5.2.1.3 Deleted.

5.2.1.4 Deleted.

5.2.1.5 Report of Test Results: The leakage rate results of Type B and C tests that meet the acceptance criteria shall be reported in the applicable -

1,ACDWR operating report. Leakage test results of Type B and C tests that fail to meet the acceptance criteria shall be reported in a separate summary that includen en analysis and interpretation of the test data, the least squares fit nr.alysis of the test data, the instrumentation error analysis, and the structural conditions 6f the containment or components, if any, which contributed to the failure in meetti,2 t!.c acceptance criteria. (

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5.2.1.6 CONTAINMENT VENTILATION ISOLATION VALVE LEAKAGE TESTS (a) Tests: The containment ventilation system dampers shall be subjected to leakage tests, in addition to the tests required by Section 5.2.1.2. The tests shall be conducted at an initial pressure of at least 52 peig.

(b) Acceptance Criteria: Excessive degradation is determined not to exist and the isolation valve (s) is considered operable if Pressure decreases by no greater than 10 psi in a ten minute test period.

(c) Corrective Action: If excessive degradation exists, the leakage path must be repaired or isolated, and retested until the criteria is set.

(d) Test Frequency: The leakage tests of the containment ventilation isolation dampers shall be conducted at least quarterly.

5.2.2 REACTOR BUILDING ISOLATION AND VACUyM RELIST VALVE OPERABILITY TESTS 5.2.2.1 The reactor building isolation system will be tested for proper operation at least once per 18 months. l 5.2.2.2 The reactor building vacuum relief valves will be tested for proper operation during each reactor shutdown for refueling but in no case at intervals greater than two years. j l

1 5-5 Amendment No. 62

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5.2.3 Tho exterior surfaces' of the LACBWR ventilation stack and the smoke stack of the conventional steam power generating station, Genon-3, adjacent to the 1.ACDWR plant shall be inspected for structural integrity at an interval no longer than five years following the initial construction inspection, and at subsequent intervals no longer than five years apart.

l 5.2.4 Deleted.

l 5.2.5 Deleted.

5.2.6 Deleted.

5.2.7 Deleted.

5.2.8 Deleted.

5.2.9 Deleted. l 5.2.10 The door seals on the containn.ent personnel and emergency airlocks will be visually inspected for degradation every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

5.2.11 The door seals on the containment personnel and emergency airlocks l will be replaced periodically in accordance with manufacturers recommendations.

5.2.12 Deleted.

5.2.13 Deleted.

5.2.14 Deleted.

5.2.15 Instrumentation shall be checked, tested and calibrated as indicated  !

in the following chart. l I

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• MINIMIN FREQ11ENCTES FOR TESTING, CALIBRATING, AND/OR CilECKING OF INSTRINENTATION l l CIIANNEl.S ACTION  : MINIM 1glFREQUENCY ___

l1. Rent: tor Water Level Calibration  : At least once per 18 months.  :

Test  : Monthly when in service.  :

Daily.  :

Check '

At least once per 18 months.

2. Area Radiation  : Calibration

• l

Monitors  :

: Test  : Quarterly.

: Check  : Daily.

Semi-annually.  :

l3. Portable Radiation  : Calibration

Detectors  :

: Check l Every two weeks.  :

4. Resetor Building  : Calibration  : At least once per 18 months.  :

Pressure  :  : '

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(Pages 5 5-10) Amendment No. 62

TABLE 1 ,

OPERATING LIMITS KEYSWITCH CHANNEL OR SENSOR SET POINT ACTION BYPASS PROVISION CONDITION

1. Reactor Water Level Water Level Safety 112" below nominal 1) Closure of Reactor One channel may be Low Channel 1 or 2 indicated level Building Steam bypassed for calibra - ,

Isolation Valve tion and testing.

and its bypass

2) Closure of Reactor Blowdown Through Decay Heat Removal Valve

3) Closure of Shutdown Condenser Condensate Drain Valve

4) Closure of Ventila-tion Inlet and f Outlet Dampers

5) Closure of Contain-C ment Offgas Vent Header Valve

6) Closure of Heating Steam Condensate Return Valve

7) Closure of Retention Tank Pump Discharge Valve a

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TABLE 1 - OPERATING LIMITS - (cont'd) ,

KEYSWITCH

- SET POINT ACTION RYPASS PROVISION CONDITION CHANNEL OR SENSOR Reactor Building Pressure < 5 psig 1) Closure of Venti- None

2. Reactor Building Transmitter 1 or 2 lation Inlet &

Pressure High Outlet Dampers

2) Closure of Containment Offgas Vent Header Valve

3) Closure of Retention Tank Pump Discharge Valve

4) Closure of Shutdown

~

Condenser Condensate Drain Valve

5) Closure of Reactor Blowdown Through Decay Heat Removal Valve

6) Closure of

}'

~J Containment High

, b Pressure Service Water Valve

7) Closure of Containment Demineralized Water Valve I

8) Closure of Contain-I ment Heating Steam Condensate Valve

< radiation levels 1) Closure of None

3. Reactor Building Radiation Monitors which correspond to ventilation inlet l Ventilation Exhaust

'l[g the limits of Specification and outlet dampers.

2) Closure of containment offgas 4.3.2.2.

,g vent header valve.

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Amendment No. 62

4

f. A Fire Brigade of at least 3 members shall be maintained on site at all times.*

g. Deleted,

b. The working hours of the Operators, the Duty Shift Supervisor, Mechanical Maintenance and Instrument & Electrical Technicians when performing duties which may affect nuclear safety, and llealth Physics Technicians, when performing radiation protection duties which may affect the safety of the public shall be limited.

In the event overtime must be used, the following restrictions shall be followod:

1. The specified personnel shall not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift turnover time.

2. The specified personnel shall not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7-day period.

3. A break of at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shall be allowed following overtime before the next scheduled shift for the specified personnel, if the above limits are exceeded.

In the event overtime must be used in excess of the above restrictions, the Plant Superintendent or his designate, must i i

authorize the deviation and the cause must be documented.

6.2.3 Deleted.

• . Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed two hours in order to accommodate unexpected absence of Fire Brigade members provided immediate action is taken to restore the Fire Brigade to within the minimum requirements. This provision does not permit any Fire Brigade position to be unmanned upon shift change due to an oncoming brigade menber being late or absent.

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