ML20205P408

Proposed Tech Specs Changes,Revising Reactor Bldg Structural Integrity Requirements

ML20205P408
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/09/1999
From:	ENTERGY OPERATIONS, INC.
To:
Shared Package
ML20205P389	List: 1CAN049901, Application for Amend to License DPR-51,revising Requirements Associated with ANO 1 Provisions for RB Testing & Insp ML20205P408
References
NUDOCS 9904200165
Download: ML20205P408 (23)

Text

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PROPOSED TECHNICAL SPECIFICATION CHANGES 1

9904200165 990409 PDR ADOCK 05000313 P

PDR

SECTION TITLE PAGE 3.14 HYDROGEN RECOMBINERS 66e 3.15 FUEL HANDLING AREA VENTILATION SYSTEM 66g 3.16 SHOCK SUPPRESSORS (SNUBBERS) 661 3.17 FIRE SUPPRESSION WATER SYSTEM 66m 3.18 FIRE SUPPRESSION SPRINKLER SYSTEMS 66n 3.19 CONTROL ROOM AND AUXILIARY CONTROL ROOM HALON SYSTEMS 66o 3.20 FIRE HOSE STATIONS 66p 3.21 FIRE BARRIERS 66q 3.22 REACTOR BUILDING PURGE FILTRATION SYSTEM 66r 3.23 REACTOR BUILDING PURGE VALVES 66t 3.24 EXPLOSIVE GAS MIXTURE 66u 3.25 RADIOACTIVE EFFLUENTS 66v 3.25.1 Radioactive Liquid Holdup Tanks 66v 3.25.2 Radioactive Gas Storage Tanks 66w 4.

SURVEILLANCE REQUIREMENTS 67 4.1 OPERATIONAL SAFETY ITEMS 67 4.2 REACTOR COOLANT SYSTEM SURVEILLANCE 76 4.3 TESTING FOLLOWING OPENING OF SYSTEM 78 4.4 REACTOR BUILDING 79 4.4.1 Reactor Building Leakage Tests 79 4.5 EMERGENCY CORE COOLING SYSTEM AND REACTOR l

j BUILDING COOLING SYSTEM PERIODIC TESTING 92 4.5.1 Emergency Core Cooling Systems 92 4.5.2 Reactor Building Cooling Systems 95 4.6 AUXILIARY ELECTRICAL SYSTEM TESTS 100 4.7 REACTOR CONTROL ROD SYSTEM TESTS 102 4.7.1 Control Rod Drive System Functional Tests 102 4.7.2 Control Rod Program Verification 104 4.8 EMERGENCY FEEDWATER PUMP TESTING 105 4.9 REACTIVITY ANOMALIES 106 4.10 CONTROL ROOM EMERGENCY AIR CONDITIONING AND ISOLATION SYSTEM SURVEILLANCE 107 4.11 PENETRATION ROOM VENTILATION SYSTEM SURVEILLANCE 109 4.12 HYDROGEN RECOMBINERS SURVEILLANCE 109b 4.13 EMERGENCY COOLING POND 110a 4.14 RADIOACTIVE MATERIALS SOURCES SURVEILLANCE 110b 4.15 AUGMENTED INSERVICE INSPECTION PROGRAM FOR HIGH ENERGY LINES OUTSIDE OF CONTAINMENT 110c i

Amendment No. M,34,M,44,44, M, M, il M,84, M,94, MG,4M,44

N LIST OF FIGURES Number Title Page 3.1.2-1 REACTOR COOLANT SYSTEM HEATUP AND COOLDOWN LIMITATIONS 20a 3.1.2-2 REACTOR COOLANT SYSTEM NORMAL OPERATION-HEATUP LIMITATIONS 20b 3.1.2-3 REACTOR COOLANT SYSTEM, NORMAL OPERATION COOLDOWN LIMITATIONS 20c 3.1.9-1 LIMITING PRESSURE VS. TEMPERATURE FOR CONTROL ROD DRIVE OPERATION WITH 100 STD CC/ LITER H O 33 2

3.2-1 BORIC ACID ADDITION TANK VOLUME AND CONCENTRATION VS. RCS AVERAGE TEMPERATURE 35a 3.5.4-1 INCORE INSTRUMENTATION SPECIFICATION AXIAL IMBALANCE INDICATION 53a 3.5.4.2 INCORE INSTRUMENTATION SPECIFICATION RADIAL FLUX TIIT INDICATION 53b 3.5.4-3 INCORE INSTRUMENTATION SPECIFICATION 53c 3.8.1 SPENT FUEL POOL ARRANGEMENT UNIT 1 59c 3.8.2 MAXIMUM BURNUP VS INITIAL ENRICHMENT FOR REGION 2 STORAGE 59d 3.24-1 HYDROGEN LIMITS FOR ANO-1 WASTE GAS SYSTEM 110bc l

4.18.1 UPPER TUBE SHEET VIEW OF SPECIAL GROUPS PER SPECIFICATION 4.18.3.a.3 110o2 5.4-1 ANO-1 FFSR LOADING PATTERN 116a Ainendment No. M,M,M,4%,444, iv 4-H,449, M4,4M

l s

3.6 REACTOR BUILDING Applicability Applies to the operability of the reactor building.

l Objective To assure reactor building operability.

l Specification 3.6.1 The reactor building shall be operable whenever all three (3) of the l

following conditions exist:

a.

Reactor coolant pressure is 300 psig or greater.

b.

Reactor coolant temperature is 200*F or greater.

c.

Nuclear fuel is in the core.

With the reactor building inoperable, restore the reactor building to operable status within one hour or be in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

3.6.2 Reactor building integrity shall be maintained when the reactor coolant system is open to the reactor building atmosphere and the requirements for a refueling shutdown are not met.

The provisions of Specification 3.0.3 are not applicable.

3.6.3 Positive reactivity insertions which would result in the reactor being suberitical by less than 1% Ak/k shall not be made by control rod motion or boron dilution whenever reactor building integrity is not in force. The provisions of Specification 3.0.3 are not applicable.

3.6.4 The reactor shall not be taken critical or remain critical if the reactor building internal pressure exceeds 3.0 psig or a vacuum of 5.5 inches Hg.

With the reactor critical, restore the containment pressure to within its limits within one hour or be in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

3.6.5 Prior to criticality following a refueling shutdown, a check shall be made to confirm that all manual reactor building isolation valves which should be closed are closed and locked, as required.

The provisions of Specification 3.0.3 are not applicable.

Amendment No. 67 54

3.6.6 If, while the reactor is critical, a reactor building isolation valve is determined to be inoperable in a position other than the closed position, the other reactor building isolation valve (except for check valves) in the line shall be tested to insure I

operability.

If the inoperable valve is not restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be brought to the cold shutdown condition within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the operable valve will bc closed.

Bases Included in reactor building operability are both the reactor building integrity as defined in Specification 1.7 and the reactor building structural integrity.

Structural integrity limitations as described in the ANO containment inspection program ensure the reactor building will be maintained comparable to the original design standards throughout the facility life span. Visual and other required examinations of tendons, anchorages and surfaces are performed periodically in accordance with station procedurec.

These procedures embody applicable requirements of the 1992 Edition with the 1992 Addenda of Subsection IWL of the ASME Boiler and Pressure Vessel Code as set forth in 10 CFR 50.55a (g) (6) (ii) (B). Any degradations reaching specific thresholds noted during inspections are evaluated by submitting a condition report and/or, within 30 days of identification, completing an engineering evaluation to determine what impact the degradation has on overall containment operability, if any.

The reactor coolant system conditions of cold shutdown assure that no steam will be formed and hence there will be no pressure buildup in the reactor building if the reactor coolant system ruptures.

The selected shutdown conditions are based on the type of activities that are being carried out and will preclude criticality in any occurrence.

The reactor building is designed for an internal pressure of 59 psig and an external pressure 3.0 psi greater than the internal pressure. The design external pressure of 3.0 psi corresponds to a margin of 0.5 psi above the differential pressure that could be developed if the building is sealed with an internal temperature of 110*F and the building is subsequently cooled to an internal temperature of less than 50*F.

When reactor building integrity is established, the limits of 10 CFR 100 will not be exceeded should the maximum hypothetical accident occur.

REFERENCE FSAR, Section 5.

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1 Amendment No. 67, 55 1

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s Bases (1)

The reactor building is designed for an internal pressure of 59 psig and a steam-air mixture temperature of 285'F.

The peak calculated reactor building pressure for the design basis loss of coolant accident, Pa, is 54 psig. The maximum allowable reactor building leakage rate, La, shall be 0.20% of containment air weight per day at Pa.

The reactor building will be periodically leakage tested in accordance with the Reactor Building Leakage Rate Testing Program.

These periodic testing I

l requirements verify the reactor building leakage rate does not exceed the l

assumptions used in the safety analysis. At s 1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 La for the combined Type B and Type C leakage, and s 0.75 La for overall Type A leakage.

At all other times between required leakage tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 La-REFERENCE (1) FSAR, Sections 5 and 13.

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l Amendment No.

+a+,+4+,+46,+66 80 Next page is 92 l

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6.12.5 Special Reports Special reports shall be submitted to the Administrator of the appropriate Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification.

a.

Deleted l

b.

Inoperable Containment Radiation Monitors, Specification 3.5.1, Table 3.5.1-1.

c.

Deleted d.

Steam Generator Tubing Surveillance - Category C-3 Results, Specification 4.18.

e.

Miscellaneous Radioactive Materials Source Leakage Tests, Specification 3.12.2.

i f.

Deleted l

g.

Deleted l

h.

Inoperable Fire Detection Instrumentation 1.

Inoperable Fire Suppression Systems j.

Degraded Auxiliary Electrical Systems, Specification 3.7.2.H.

k.

Inoperable Reactor Vessel Level Monitoring Systems, Table 3.5.1-1 1.

Inoperable Hot Leg Level Measurement Systems, Table 3.5.1-1 m.

Inoperable Main Steam Line Radiation Monitors, Specification 3.5.1, Table 3.5.1-1.

Amendment. No. 88,448,4% MB, M3, 146a 444

A 4

t MARKUP OF CURRENT ANO-1 TECHNICAL SPECIFICATIONS (FOR INFO ONLY)

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SECTION TITLE PAGE l

3.14 HYDROGEN RECOMBINERS 66e

)

3.15 FUEL HANDLING AREA VENTILATION SYSTEM 66g 3.16 SHOCK SUPPRESSORS (SNUBBERS) 661 3.17 FIRE SUPPRESSION WATER SYSTEM 66m 3.18 FIRE SUPPRESSION SPRINKLER SYSTEMS 66n 3.19 CONTROL ROOM AND AUXILIARY CONTROL ROOM HALON SYSTEMS 66o 3.20 FIRE HOSE STATIONS 66p 3.21 FIRE BARRIERS 66q 3

3.22 REACTOR BUILDING PURGE FILTRATION SYSTEM 66r 3.23 REACTOR BUILDING PURGE VALVES 66t 3.24 EXPLOSIVE GAS MIXTURE 66u 3.25 RADIOACTIVE EFFLUENTS 66v 3.25.1 Radioactive Liquid Holdup Tanks 66v 3.25.2 Radioactive Gas Storage Tanks 66w 4.

SURVEILLANCE REQUIREMENTS 67 4.1 OPERATIONAL SAFETY ITEMS 67 4.2 REACTOR COOLANT SYSTEM SURVEILLANCE 76 4.3 TESTING FOLLOWING OPENING OF SYSTEM 78 4.4 REACTOR BUILDING 79 4.4.1 Reactor Building Leakage Tests 79 4.4.2 Structural I.tegrity 96 l

4.5 EMERGENCY CORE COOLING SYSTEM AND REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 92 4.5.1 Emergency Core Cooling Systems 92 4.5.2 Reactor Building Cooling Systems 95 4.6 AUXILIARY ELECTRICAL SYSTEM TESTS 100 4.7 REACTOR CONTROL ROD SYSTEM TESTS 102 4.7.1 Control Rod Drive System Functional Tests 102 4.7.2 Control Rod Program Verification 104 4.8 EMERGENCY FEEDWATER PUMP TESTING 105 4.9 REACTIVITY ANOMALIES 106 4.10 CONTROL ROOM EMERGENCY AIR CONDITIONING AND ISOLATION SYSTEM SURVEILLANCE 107 4.11 PENETRATION ROOM VENTILATION SYSTEM SURVEILLANCE 109 4.12 HYDROGEN RECOMBINERS SURVEILLANCE 109b 4.13 EMERGENCY COOLING POND 110a 4.14 RADIOACTIVE MATERIALS SOURCES SURVEILLANCE 110b 4.15 AUGMENTED INSERVICE INSPECTION PROGRAM FOR HIGH ENERGY LINES OUTSIDE OF CONTAINMENT 110c Amendment No. M, B4, M, M, 44, 64, 66, 11 64, 84, %%, %, HB,42, M3

I LIST OF FIGURES Number Title Page l

3.1.2-1 REACTOR COOLANT SYSTEM HEATUP AND COOLDOWN LIMITATIONS 20a 3.1.2-2 REACTOR COOLANT SYSTEM NORMAL OPERATION-HEATUP i

LIMITATIONS 20b j

3.1.2-3 REACTOR COOLANT SYSTEM, NORMAL OPERATION COOLDOWN LIMITATIONS 20e 3.1.9-1 LIMITING PRESSURE VS. TEMPERATURE FOR CONTROL ROD DRIVE OPERATION WITH 100 STD CC/ LITER H O 33 6

2 3.2-1 BORIC ACID ADDITION TANK VOLUME AND CONCENTRATION VS. RCS AVERAGE TEMPERATURE 35a 3.5.4-1 INCORE INSTRUMENTATION SPECIFICATION AXIAL IMBALANCE j

INDICATION 53a l

3.5.4.2 INCORE INSTRUMENTATION SPECIFICATION RADIAL FLUX TILT INDICATION 53b 3.5.4-3 INCORE INSTRUMENTATION SPECIFICATION 53c l

3.8.1 SPENT FUEL POOL ARRANGEMENT UNIT 1 59c 3.8.2 MAXIMUM BURNUP VS INITIAL ENRICHMENT FOR REGION 2 STORAGE 59d 3.24-1 HYDROGEN LIMITS FOR ANO-1 WASTE GAS SYSTEM 110be j

t.2 1 NORMALI2EO L!! TOFF FORCE "000 TENDON S5b 4r4ve-2 NORM ^.LIZED

'"OEF FORCE DOME TENDONS 95e j

i 1.2-2 NORMALIZE 0 LIFTOFF FonCE VESTICAL TEMOONS 95d j

4.18.1 UPPER TUBE SHEET VIEW OF SPECIAL GROUPS PER SPECIFICATION 4.18.3.a.3 110o2 5.4-1 ANO-1 FFSR LOADING PATTERN 116a Amendment No. M,M,93,4M,444, iv 4-M,4M,4M, W3

's 3.6 REACTOR BUILDING Applicability Applies to the i.tegrit-y-opsIAhilitLof the reactor building.

l Objective To assure reactor building 4*tegel+yoperability.

l Specification 3.6.1 Ihe_IReactor building htegrity shall be mahteined-operable whenever all l three (3) of the following conditions exist:

a.

Reactor coolant pressure is 300 psig or greater.

b.

Reactor coolant temperature is 200*F or greater.

c.

Nuclear fuel is in the core.

Without !;he_ reactor building heegel+yinoperable, restore the-htcg rity reactor buildino to operab_le status within one hour or be in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

3.6.2 Reactor building integrity shall be maintained when the reactor coolant system is open to the reactor building atmosphere and the requirements for a refueling shutdown are not met.

The provisions of Specification 3.0.3 are not applicable.

3.6.3 Positive reactivity insertions which would result in the reactor being suberitical by less than 1% Ak/k shall not be made by control rod motion or boron dilution whenever reactor building integrity is not in force. The provisions of Specification 3.0.3 are not applicable.

3.6.4 The reactor shall not be taken critical or remain critical if the reactor building internal pressure exceeds 3.0 psig or a vacuum of 5.5 inches Hg.

With the reactor critical, restore the containment pressure to within its limits within one hour or be in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

3.6.5 Prior to criticality following a refueling shutdown, a check shall be made to confirm that all manual reactor building isolation valves which should be closed are closed and locked, as required.

The provisions of Specification 3.0.3 are not applicable.

Amendment No. 64, 54

3.6.6 I f, while the reactor is critical, a reactor building isolation valve is determined to be inoperable in a position other than the closed position, the other reactor building isolation valve (except for check valves) in the line shall be tested to insure operability.

If the inoperable valve is not restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be brought to the cold shutdown condition within an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the operable valve will be closed.

Bases locluded in reactor _ building _pperability are both the reactor buildin_g integrity as defined in Specification 1.7 and the reactor building structural integr13ya St ructu ral_intstgrit y limita31gns as described in the ANO containment inspect _1gn procram ensure the reactor buildina will be maintained compaIable to the CI1gina_1_._d_e s igtl_sta nda rd s throuchout the facility life span.

Visual and other required examinations of tendons, anchorages and surfaces are performed periodically in accordance with station procedures. These_ procedures embody gpplicable requirgme_nts o_fLthe 1992 Edition with the 1992 Addenda of_ Subsection IWL_o.f the ASME_ Roller and Pressure Vessel Code as set forth in 10 CFR 50.55a (c) (6) (ii) (B).

Any degradations reaching specific thresholds noted during inspections are evaluated by subm_it_ tine a condition report __aILdign _within 30 day.s of identification, compl_etine an encineerino evaluation to detetmine what irnpact the degradation has on overall containment ope _tability, if any.

The reactor coolant system conditions of cold shutdown assure that no steam will be formed and hence there will be no pressure buildup in the reactor building if the reactor coolant system ruptures.

The selected shutdown conditions are based on the type of activities that are being carried out and will preclude criticality in any occurrence.

The reactor building is designed for an internal pressure of 59 psig and an external pressure 3.0 psi greater than the internal pressure. The design external pressure of 3.0 psi corresponds to a margin of 0.5 psi above the j

differential pressure that could be developed if the building is sealed with an internal temperature of 110*F and the building is subsequently cooled to an internal temperature of less than 50*F.

When reactor building integrity is established, the limits of 10 CFR 100 will not be exceeded should the maximum hypothetical accident occur.

REFERENCE FSAR, Section 5.

l 1

Amendment No. 64, 55 l

I t

Bases (1)

The reactor building is designed for an internal pressure of 59 psig and a steam-air mixture temperature of 285'F.

The peak calculated reactor building pressure for the design basis loss of coolant accident, Pa, is 54 psig.

The maximum allowable reactor building leakage rate, La, shall be 0.20% of containment air weight per day at Pa-1 The r(*

• tor building will be periodically leakage tested in accordance with 1

the Reactor Building Leakage Rate Testing Program. These periodic testing requirements verify the reactor building leakage rate does not exceed the assumptions used in the safety analysis. At s 1.0 La the offaite dose con-ruences are bounded by the assumptions of the safety analysis.

During the tirst unit startup followir.g testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 La for the combined Type B and Type C leakage, and s 0.75 La for overall Type A leakage. At all other times between required leakage tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 La-l l

REFERENCE l

l (1) FSAR, Sections 5 and 13.

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1 r-6.12.5 Special Reports Special reports shall be submitted to the Administrator of the appropriate Regional Office within the time period specified for each report.

These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification.

a.

Tenden Surveillance, Sp ification

'.'.2.2 Deleted l

b.

Inoperable Containment Radiation Monitors, Specification 3.5.1, Table 3.5.1-1.

c.

Deleted d.

Steam Generator Tubing Surveillance - Category C-3 Results, Specification 4.18.

e.

Miscellaneous Radioactive Materials Source Leakage Tests, Specification 3.12.2.

f.

Deleted g.

Deleted h.

Inoperable Fire Detection Instrumentation i.

Inoperable Fire Suppression Systems j.

Degraded Auxiliary Electrical Systems, Specification 3.7.2.H.

4 k.

Inoperable Reactor Vessel Level Monitoring Systems, Table 3.5.1-1 1.

Inoperable Hot Leg Level Measurement Systena, Table 3.5.1-1 m.

Inoperable Main Steam Line Radiation Monitors, specification 3.5.1, Table 3.5.1-1.

Amendment No. 44,444,464,444,M3, 146a 444