ML20034H717

Amend 151 to License DPR-40,revising TS 2.6 & Definition of Ci to Reflect Combustion Engineering STS for Ci & Pal

ML20034H717
Person / Time
Site:	Fort Calhoun
Issue date:	03/16/1993
From:	Hubbard G Office of Nuclear Reactor Regulation
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ML20034H716	List: ML20034H717
References
NUDOCS 9303190309
Download: ML20034H717 (19)

Text

oag g^o UNITED STATES

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NUCLEAR REGULATORY COMMISSION E(b oy WASHINGTON,0 C.20555

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OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 l

FORT CALHOUN STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.151 License No. DPR-40 1.

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

A.

The application for amendment by the Omaha Public Power District (the licensee) dated February 13, 1992, as supplemented January 8 and January 13, 1993, 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 Commission; 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; D.

The issuance of this license 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.

P

' 2.

Accordingly, Facility Operating License No. DPR-40 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No.

DPR-40 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised throu the license. gh Amendment No.151, are hereby incorporated in The licensee shall operate the facility in accordance with the Technical Specifications.

3.

The license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION oft

<v 7t+C George T. Hubbard, Acting Director Project Directorate IV-I Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

March 16, 1993 3

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...m.

s ATTACHMENT TO LICENSE AMENDMENT NO.151 i

FACILITY OPERATING LICENSE NO. DPE,; j i

DOCKET NO. 50-285 I

i Revise Appendix "A" Techiiical Specifications as indicated below. The revised pages are identified by amendment number and contain vertical lines indicating

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the area of change.

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REMOVE PAGES INSERT PAGES 5

5 2-30 2-30 i

2-31 2-31 2-31a 2-31a 3

3-37 3-37 i

3-38 3-38 3-39 3-39 3-40 3-40 3-42 3-42 3-43 3-43 i

3-45 3-45 i

3-47 3-47 3-48 3-48 3-49 3-49 3-50 3-50 3-51 3-51 I

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I DEFINITIONS MISCELLANEOUS DEFINITIONS Operable - Onerability A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, 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 their related support function (s).

In Operation A system or component is in operation if it is performing its design function.

CE A's All full length shutdown and regulating control rods.

Non-trionable (NT) CEA's CEA's which are non-trippable.

Containment Integrity Containment integrity is defined to exist when all of the following are met:

(1)

All nonautomatic containment isolation valves which are not required to be open during accident conditions and blind flanges, except for valves that are open under administrative control as permitted by Specification 2.6(1)a, are closed.

(2)

The equipment hatch is properly closed and sealed.

(3)

The personnel air lock satisfies Specification 2.6(1)b.

l (4)

All automatic containment isolation valves are operable, locked closed, or deactivated and secured in their closed position (or isolated by locked closed valves or blind flanges as permitted by a limiting condition for operation).

_(5)

Tine uncontrolled containment leakage satisfies Specification 3.5, and I

(6)

The sealing mechanism associated with each penetration (e.g., welds, bellows or O-rings) is operable.

5 Amendment No. g09,

2.0 LIMITING CONDITIONS FOR OPERATION 2.6-fontainment System Applicability Applies to the reactor containment system.

Objective To assure the integrity of the reactor containment system.

Specifications (1)

Containment Inteerity a.

Containment integrity shall not be violated unless the reactor is in a cold or refueling shutdown condition. Without containment integrity, restore containment integrity within one hour or be in at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, in at least suberitical and <300*F 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 />. Normally locked or sealed-closed valves (except for PCV-742A/B/C/D) may be opened intermittently under administrative control without constituting a violation of containment integrity.

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b.

The personnel air lock shall be operable unless the reactor is in a cold or refueling shutdown condition. Both doors shall be closed except when the air lock is being used for normal transit, then at least one air lock door shall be closed. The entire air lock assembly leakage rate shall be in accordance with Specification 3.5(4).

(i).

With one personnel air lock door inoperable.

a.

Maintain at least the operable air lock door closed and either restore the inoperable air lock door to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or lock the operable air lock door closed.

Entry and exit is permissible to perform repairs of the affected air lock components without constituting a violation of containment integrity, b.

Operation may then continue until performance of the next required entire air lock assembly leakage test provided that the operable air lock door is verified to be locked closed at least once per 31 days.

c.

Othenvise, be in at least hot shutdown 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 />.

d.

Entry into another operational mode or specified condition is allowed if the provisions stated in 2.6(1)b.(i)a. above are -

m et.

(ii).

With the personnel air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least hot shutdown 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 />.

j 2-30 Amendment No. g138,

5 2.0 LIMITING CONDITIONS FOR OPERATION 2.6 Containment System (Continued) c.

Containment integrity shall not be violated when the reactor vessel head l

is removed if the boron concentration is less than refueling concentration.

d.

Except for testing one CEDM at a time, positive reactivity changes shall l

not be made by CEA motion or boron dilution unless the containment integrity is intact.

I e.

The containment purge isolation valves will be locked closed unless the reactor is in a cold or refueling shutdown condition.

s (2)

Internal Pressure i

The intemal pressure shall not exceed 3 psig (except for containment leak rate tests).

(3)

Hydrogen Purge System a.

Minimum Reauirements The reactor shall not be made critical unless all of the following requirements are met:

i 1.

The containment isolation valves VA-280 and VA-289 shall be locked closed.

Opening of these valves intermittently under administrative control is not allowed.

2.

VA-80A and VA-80B with associated valves and piping to include VA-82 filters, are operable.

b.

Modification of Minimum Recuirements i

i After the reactor has been made critical, the minimum requirements may be modified to allow either or both of the following statements (i,ii) to be applicable at any one time. If the operability of the component (s) is not -

restored to meet the minimum requirements within the time specified 1

below, the reactor shall be placed in a hot shutdown condition within six

{

hours.

(i)

One of the hydrogen purge fans, VA-80A or VA-80B, with associated valves and piping, may be inoperable provided the fan l

is restored to operable status within 30 days.

(ii)

The hydrogen purge filter system, VA-82, may be inoperable provided the system is restored to operable statu.e within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

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l 2-31 Amendment No. 68dM, 151

2.0 LIMITING CONDITIONS FOR OPFRATION 2.6 Containment System (Continued)

BaEis The reactor coolant system conditions of cold shutdown assure that no steam will be formed and, hence, there would be no pressure buildup in the containment if the reactor coolant system ruptures. '1he shutdown margins are selected based on the type of activities that are being carried out. The refueling boron concentration provides a shutdown margin which precludes criticality under any circumstances. Each CEDM must be tested and some have two CEA's attached.

1 Regarding internal pressure limitations, the containment design pressure of 60 psig would not be exceeded if the internal pressure before a major loss-of-coolant accident were as much as 3 psig.m The opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment. Operation of the purge isolation valves is prevented during normal operations due to the size of the valves (42 inches) and a concern about their-ability to close against the differential pressure that could result from a LOCA or MSLB.-

Specification 2.6(1)a applies when both doors of the PAL are declared inoperable, or the entire air lock assembly leakage exceeds the requirements of Specification 3.5(4).

Specification 2.6(1)b(ii) applies when mechanisms other than a door, such as the inner -

door equalizing valve, are declared inoperable.

The Hydrogen Purge System is required to be operable in order to control the quantity of combustible gases in containment in a post-LOCA condition.* The containment integrity will be protected by ensuring the penetration valves VA-280 and VA 289 are

" locked closed" while HCV-881 and HCV-882 are normally closed during power operation. The applicable surveillance testing requirements of Table 3-5 will ensure that the system is capable of performing its design function. The blowers (VA-80A and VA-80B), associated valves, and piping are single faih.re proof, have been designed as a -

Seismic Class 1 System, and are redundant to the VA-82 filter header. VA-80A or VA-80B is capable of providing sufficient hydrogen removal capabilities as required by the USAR to prevent the hydrogen concentration inside of containment from exceeding the 4% flammability limit.* Electrical Equipment qualification was not required as the radiation doses in the area of the Hydrogen Purge System equipment were below the minimum requirements.")

VA-80A or VA-80B with the associated valves and piping may be inoperable for 30 days. The redundancy of the blowers allows one blower with associated valves and piping to be removed from operation while the other train has the capability to provide 100% hydrogen control.

References 1

(1)

USAR, Section 14.16; Figure 14.16-2 (2)

Regulatory Guide 1.7 (1971)

(3)

USAR, Section 14.17 (4)

Engineering Study 86-10, Calculation 53 i

2-31a Amendment No.g-

3.0 SURVEILLANCE REOUTREMENTS 3.5 Containment Test Applicability Applies to containment leakage and structural integdty.

Obiective

[

To verify that the:

(1) locked closed manual containment isolation valves are closed and locked, (2) potential leakage from contr.inment is within acceptable limits, and (3) structural performance of all important compohents in the containment prestressing system is acceptable.

Specifications (1)

Prior to the reactor going critical after a refueling outage, and at least once per 31 days thereafter, an administrative check will be made to confirm that all

" locked closed" manual containment isolation valves, except for valves that are open under administrative control as permitted by Specification 2.6(1)a, are closed and locked. Valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed or otherwise secured in the closed position shall be verified closed during each cold shutdown except that -

such verification need not be performed more often than once per 92 days.

(2)

Containment Buildine Leak Rate Tests l_

Tests shall be conducted to assure that leakage of the primary reactor containment and associated systems is maintained within allowable leakage rate limits.

Periodic surveillance shall be performed to assure proper maintenance and leak repair of the containment structure and penetrations during the plant's operating life.

Definitions of terms used in the leak rate testing specifications:

Leakage Rate - for test purposes is that leakage of containment air which occurs in a unit of time. Stated as a percentage of weight of the original content of containment air at the leakage rate test pressure that escapes to the outside atmosphere during a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> test pedod.

Maximum Allowable Leakage Rate (L)- the design basis leakage rate of 0.1%_

by weight of the containment atmosphere per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at a pressure of 60 psig.

Overall Integrated leakage Rag - that leakage rate which is obtained from a -

summation of leakage through all potential leakage paths including containment welds, valves, fittings, and components which penetrate containment.

Acceptable Criteria - the standard against which test results are to be compared for establishing the functional acceptability of the containment as a leakage

'l limiting boundary.

3-37 Amendment No. g

3.0 SURVEILLANCE REOUIREMENTS

~.

3.5 Containment Tests (Continued)

(3)

Integated 12ak Rate Test (Type A Test) l a.

Introduction Type A tests are intended to measure the reactor containment overall integrated leakage rate at periodic intervals, b.

Pretest Recuirements A general inspection of the accessible interior and exterior surfaces of the containment structures and components shall be performed prior to any Type A test to uncover any evidence of structural deterioration which may affect either the containment structural integrity or leak-tightness. If there is evidence of stmetural deterioration, the Type A tests shall not be performed until corrective action is taken in accordance with repair procedures, non-destructive examinations, and tests as specified in the applicable code specified in 10 CFR Part 50.55a at the commencement of repair work. Such structural deterioration and corrective actions taken shall be reported as part of the Type A test report.

During the period between the initiation of the containment inspection and performance of the Type A test, no repairs or adjustmenis shall be made so that the containment can be tested in as close to the "as is" condition as practical. During the period between the completion of one Type A test and the initiation of the containment inspection for the subsequent Type A test, repairs or adjustments shall be made to components whose leaksge exceeds that specified in the Technical Specifications as soon as practical after identification.

This requirement is interpreted not to preclude performance of Type B and Type C testing ano required repairs prior to initiation of the containment inspection and the performance of the Type A test.

If during a Type A test, potentially excessive leakage paths are identified which interfere with satisfactory completion of the test, or which result in the Type A test not meeting the acceptance criteria, the Type A test shall be temporarily suspended. Thereafter, repairs and/or adjustments 13 i

equipment shall be made and the Type A test resumed. The corrective action taken, the change in leakage rate resulting from the repairs'and overall integrated leakage determined from the Type A and local leak rate tests shall be included in a repon submitted to the Commission.

Closure of containment isolation valves for the Type A test shall be accomplished by normal operation and without any preliminary exercising or adjustments (e.g., no tightening of valve after closure by valve monitor).

Repairs of maloperating or leaking valves shall be made necessary.

Information on any valve closure malfunction or valve i

leakage that requires corrective action before the test, shall be included in the Type A Izak Test Repon submitted to the Commission.

t 3-38 Amendment No. 9h 151

3.0 SURVEILLANCE REOUIREMENTS 3.5 Containment Tests (Continued)

I The containment test conditions shall stabilize for a period of approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to the start of the leakage rate test.

Those portions of the fluid systems that are part of the reactor coolant pressure boundary and are open directly to the containment atmosphere under post-accident conditions and become an extension of the boundary of the containment shall be opened or vented to the containment atmosphere prior to and during the test. Portions of closed systems inside containment that penetrate containment and rupture as a result of a loss of coolant accident shall be vented to the containment atmosphere. All vented systems shall be drained of water or other fluids to the extent.

necessary to assure exposure of the system containment isolation valves to containment air test pressure and to assure they will be subjected to the post-accident differential pressure. Systems that are required to maintain the plant in a safe condition during the test shall be operable in their normal mode, and need not be vented. Systems that are normally filled with water and operating under post-accident conditions, such as the containment heat removal system and the component cooling water system, need not be vented. However, the containment isolation valves in the systems defined in this section shall be tested in accordance with Section 3.5(5). The measured leakage rate from these tests shall be l

reported to the Commission.

c.

Iest Methods All Type A tests shall be conducted in accordance with the provisions of 10 CFR Part 50, Appendix J.

I The accuracy of any Test A shall be verified by a supplemental test. The supplemental test method selected shall be conducted for sufficient duration to establish accurately the change in leakage rate between the Type A test and the supplemental Type A test.

Results from the supplemental test are acceptable provided the difference between the supplemental test data and the Type A test data is within 0.25 L,. If results are not within 0.25 L., the reason shall be determined, corrective action taken, and a successful supplemental test performed.

Test leakage rates shall be calculated using absolute values corrected for instrument error, d.

Accentance Criteria The maximum allowable leakage rate shall not exceed 0.1 %.

3-39 Amendment No. 957 151

I 3.0 SURVEILLANCE REOUIREMENTS 3.5 Containment Tests (Continued)

The total measured leakage rate at a pressure of 60 psig shall be less than 0.75 L,.

If local leakage measurements are taken to effect repairs in order to meet 0.75 L, acceptance criteria, these measurements shall be taken at a pressure of 60 psig.

If two consecutive Type A tests fail to meet the acceptance criteria, notwithstanding the requirements of the testing frequency, a Type A test shall be performed at each refueling outage or approximately every 18 months, whichever occurs first, until two consecutive Type A tests meet the acceptance criteria, after which time the normal testing frequency schedule may be resumed.

c.

Testine Frecuency A set of three Type A tests shall be performed, at approximately equal intervals during each 10 year service period. The third test of each set shall be conducted when the plant is shutdown for the 10-year in-service inspections.

The performance of Type A tests shall be limited to periods when the plant facility is non-operational and secured in the shutdown condition under administrative control and in accordance with the safety procedures defined in the license.

(4)

Containment Penetrations Leak Rate Tests (Tvne B Tests) l a.

Introduction Type B tests are intended to detect local leaks and to measure leakage across each pressure-containing or leakage limiting boundary for the containment penetrations.

b.

Test Methods Type B tests shall be performed by local pneumatic pressurization of the containment penetrations, either individually or in groups, at a pressure of 60 psig.

i i

Examination shall be performed by halide leak-detection method or by other equivalent test methods such as measurement of the rate of makeup j

required to maintain the test volume at 60 psig.

j 3-40 Amendment No. 9h 151

1 7

3.0 SURVEILLANCE REOUIREMENTS i

3.5 Containment Tests (Continued) i A-1 B-9 D4' F-2 E-HCV 383-3A A2 B-10 D-7 F E-HCV-383-3B A4 B-11 D-8 F-5 E-HCV-383-4A A-5 C-1 D-9 F4 E-HCV-383-4E A4 C-2 D-10 F-7 A-7 C-4 D-11 F-8 A-8 C-5 E-1 F-9 A-9 C4 E-2 F-10 A-10 C-7 E-4 F-11 q

A-11 C-8 E-5 G-1 B-1 C9 E4 G-2 B-2 C 10 E G-3 B-4 C-11 E-8 G-4 B-5 D-1 E-9 H-1 B4 D-2 E-10 H-2 B-7 D-4 E-11 H-3 B-8 D-5 F-1 H-4 (5)

Containment Isolation Valves 12ak Rate Tests (Tyne C Testst l

a.

Introduction Type C tests are intended to measure containment isolation valve leakage rates.

b.

Test Methods Type C tests shall be performed by local pressurization with air or nitrogen at a pressure of 60 psig. The pressure shall be applied in the same direction as that when the valve would be required to perform its safety function, unless it can be determined that the results from the tests for a pressure applied in a different direction will provide equivalent or more conservative results. Each valve to be tested shall be closed by normal operation and without any preliminary exercising or adjustments-(e.g., no tightening of valve after closure by valve motor).

c.

Acceptance Criteria The combined leakage rate of all penetrations and valves subject to Type B and Type C tests shall be less than or equal to 0.6 L,.

For the purge isolation valve tests, the measured purge valve leakage rate shall be substituted for the purge valve leakage rate from the last complete Type B and C test and the total leak rate recomputed.

Leakage of the containment air purge-isolation valves shall not exceed -

18,000 standard cubic centimeters per minute (SCCM). - If the leakage rate is determined to_be greater than 18,000 SCCM, repairs shall be initiated immediately in order to meet this acceptance criterion.

A 3-42 Amendment No. 6P45, 151-

i i

3.0 SURVETLLANCE REOUIREMENTS 3.5 Containment Tests (Continued)

If at any time it is determined that a leakage rate is greater than 0.6 L.,

repairs shall be initiated immediately. If repairs are not completed and conformance to the acceptance criteria is not demonstrated 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 shut down and depressurized until repairs are completed and the local leakage meets this acceptance criteria.

d.

Testing Frecuency Type C tests shall be performed during each refueling outage, or other convenient intervals, but in no case at intervals grea*.er than 2 years. The containment purge isolation valves shall also be leakage tested prior to

]

bringing the reactor out of each cold or refueling shutdown but in no case 1

at intervals greater than nine months. If the purge valves are opened during cold or refueling shutdown, the leak test shall be performed after the purge valves are closed for the last time.

e.

Penetrations to be Testedm M-2 M-31 M-52 IA-3092 M-7 M-38 M-53 IA-3093 M-8 M-39 M-57 IA-3094 M-11 M-40 M-58 M-14 M-42 M-69 M-15 M-43 M-73 M-18 M-44 M-74 M-19 M-45 M-79 M-20 M-46 M-80 M-22 M-47 M-87 i

M-24 M-48 M-88 M-25 M-50 M-HCV-383-3 M-30 M-51 M-HCV-383-4 (6)

Special Testine Reauirements

[

i Any major modification or replacement of a component which is part of the i

containment boundary shall be followed by either Type A, Type B, or Type C tests as applicable for the area affected by the modification and shall meet the j

applicable acceptance criteria. Minor modifications, or replacements, performed j

directly prior to the conduct of a scheduled Type A test do not require a separate -

test.

1 (7)

Report on Test Results l

leak rate tests shall be the subject of a summary technical report submitted to the Commission approximately 3 months after the conduct of each test. The report shall be titled " Reactor Containment Building Integrated Leak Rate Test."

l 3-43 Amendment No. %

151

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3.0 SURVEILLANCE REOUIREMENTS 3.5 Containment Tests (Continued) l l

l (8)

Surveillance for Prestressine System a.

Sample Selection The 210 dome tendons and 616 helical wall tendons shall be periodically inspected for symptoms of material deterioration or prestressing force reduction.

Inspections shall be performed on four dome tendons, one from each layer and the control dome tendon, and ten helical wall tendons, five of each orientation including one control tendon in each orientation.

The tendons to be inspected shall be randomly selected from the tendons which have not been tested in previous surveillances, except for the control tendons which shall be included in each surveillance-sample selection to develop a historical trend in order to correlate the obsenred data.

b.

Visual Inspection The following visual inspections shall be performed:

(i)

The exterior surface of the containment shall be visually examined to detect areas of large spall, severe scaling, D-cracking in areas of 25 square feet or more, grease leabge, and other significant structural deterioration or disintegration.

(ii)

For each surveillance tendon, selected in accordance with 3.5(8)a., the tendon anchorage assembly hardware shall be visually inspected for signs of abnormal material behavior or wear.

P (iii)

The concrete surrounding the visually inspected tendon anchorages shall be visually inspected for signs of significant structural deterioration.

(iv)

The bottom grease caps of all' helical wall tendons shall be visually inspected to detect grease leakage or grease cap deformations. Removal i

of the grease caps is not necessary for this inspection.

c.

Prestress Monitorine Tests 6

Liftoff tests shall be performed on each tendon selected in accordance with 3.5(8)a. to monitor prestress. Additionally, the tests shall include the following:

l l

i 3-45 Amendment No. 95,97,139, 1

151 1

3.0 SURVEILLANCE REOUIREMENTS 3.5 Containment Tests (Continued)

The tendons detensioned in accordance with 3.5(8)c.(i) may be the tendons l

from which the sample wires are removed. The control tendons shall NOT be included as tendons to be detensioned or have wires removed.

In addition, all wires found to be broker, shall be removed for tensile testing ar,d visual examination.

e.

Instection of Filler Grease A sample of sheathing filler grease from each of the sample tendons shall be taken and analyzed according to the following national standards:

(i)

To determine water content, ASTM D95, " Standard Test Methods for Water in Petroleum Products and Bituminous Materials by Distillation."

(ii)

To determine reserve alkalinity, ASTM D974, " Standard Test Method for Acid and Base Number by Color-Indicator Titration."

(iii)

To determine the concentration of water soluble chlorides, ASTM D512,

" Standard Test Methods for Chloride Ion in Water."

(iv)

To determine the concentration of water soluble nitrates, ASTM D3867,

" Standard Test Methods for Nitrite-Nitrate in Water."

(v)

To determine the concentration of water soluble sulfides, APHA 4500-S.

2 D.

" Methylene Blue Method," Standard Methods for Examination of Waki and Waste Water. Seventeenth Edition.

In addition to these tests, the amount of filler grease removed from and replaced into each surveillance tendon shall be recorded and compared to assess grease leakage within the containment structure.

f.

Acceptance Criteria (i)

No evidence of significant structural deterioration of the concrete inspected in accordance with 3.5(8)b.(i) and 3.5(8)b.(iii) which may affect l

the structural integrity of the containment structure can be detected.

3-47 Amendment No. 95,97,139 151

3.0 SURVEILLANCE REOUIREMENTS 3.5 Containment Tests (Continued)

Significant structural deterioration is defined as measurable structural deterioration which, when compared with past inspections, shows strong evidence of an increase of structural deterioration which could affect the Containment's structural integrity. Evidence of cosmetic or superficial deterioration, unless determined by sound engineering judgement to be significant, is not considered to be significant structural deterioration.

No evidence of significant material degradation or corrosion of tendon-anchorage hardware can be detected.

If any grease leakage is detected during visual examination of the containment exterior surface, an investigation shall be made to determine the extent of potential reduction of Containment structural integrity. An investigation shall also be made to determine which tendons could have lost the grease and whether the grease loss has adversely affected their corrosion protection.

(ii)

The prestressing force measured for each tendon liftoff tested in accordance with 3.5(8)c. shall be compared with the limits predicted by l

USAR Fig 5.10-3. If the measured prestressing force of a selected tendon is greater than the prescribed lower limit, the tendon is acceptable.

If the measured prestressing force of a selected tendon is less than the prescribed lower limit but greater than or equal to 95 % of the prescribed l

lower limit, the tendon shall be tensioned to a prestress value greater than the prescribed lower limit but less than 742 kips. After increasing the tendon's prestress the tendon will be considered acceptable.

3-48 Amendment No. 95,97,139 151

3.0-SURVEILLANCE REOUIREMENTS 3.5 Containment Tests (Continued)

If the measured prestressing force of a selected tendon is less than 95%

of the prescribed lower limit but greater than or equal to 90% of the prescribed lower limit, two additional tendons, one on each side of the i

first tendon, shall be liftoff tested. If the prestressing forces of each of the second and third tendons are greater than 95% of the prescribed lower limit, all three tendons shall be tensioned to greater than the prescribed lower limit, but less than 742 kips.

After increasing the tendons' prestress, the tendons will be considered acceptable. If the prestressing force of either the second or third tendons is less than 95% of the prescribed lower limit, liftoff tests shall be performed on additional tendons to determine the cause and extent of such occurrence. This occurrence shall be considered reportable per 3.5(8)g. If the measured l

prestressing force of a selected tendon is less than 90% of the prescribed lower limit, the defective tendon shall be fully inspected to determine the cause and extent of such occurrence. This occurrence shall be considered reportable per 3.5(8)g.

l If the average prestressing force of all measured tendons of a group (corrected for average condition) is found to be less than the prescribed lower limit, an investigation shall be performed to determine the cause and extent of such an occurrence. Such an occurrence shall be considered reportable per 3.5(8)g.

l If from consecutive surveillances the average measured prestressing force of a tendon group trends at a rate which would indicate that the loss of prestress would make the average prestress of the group of tendons less i

than the prescribed lower limit before the next surveillance, additional liftoff tests shall be performed to determine the cause and extent of such occurrence.

Such an occurrence shall be considered reportable per 3.5(8)g.

l (iii)

If during the detensioning and retensioning of tendons in accordance with 3,5(8)c., the elongation corresponding to a specific load differs by more

[

than 10% from that recorded during installation of the tendons, an investigation shall be made to ensure that the difference is not related to wire failures or slippage of wires in anchorages. A difference of more than 10% shall be considered reportable per 3.5(8)g.

l-3 1

f 3-49 Amendment No. 95 97,139, 15i

~

3.0 SURVEILLANCE REOUIREMENTS 3.5

' containment Tests (Continued)

(iv)

The minimum acceptable ultimate tensile strength of the wire samples to be tensile tested shall be 240,000 psi with a minimum elongation of 4% in accordance with ASTM A421-65 forType BA wire. Failure in the tensile test at strength or elongation values less than those specified shall be considered reportable per 3.5(8)g..

l Other conditions which indicate corrosion found by visual examination of the wire shall be considered reportable per 3.5(8)g.

l (v)

Results of the laboratory tests and examinations of the filler grease will be considered acceptable if the following conditions are met:

(a)

Water content i 10% by weight (b)

Chlorides i 10 ppm (c)

Nitrates i 10 ppm (d)

Sulfides i 10 ppm (e)

Reserve alkalinity

>0 (Base numbers)

(f)

The difference between the amount of grease injected into a tendon to replace the amount which was removed during inspection shall not exceed 5% of the net tendon sheath (duct) volume when injected at the original installation pressure.

(g)

The lack of the presence of any free water.

The failure to meet any of the above conditions for the filler grease shall be considered reportable per 3.5(8)g.

I g.

Corrective Action and Ret >ortine l

If the above acceptance criteria are not met, an immediate investigation shall be made to determine the cause(s) and extent of the nonconformance to the criteria, and the results shall be reported to the Commission within 90 days via a special report in accordance with Technical Specification 5.9.3.

3-50 Amendment No. 9468;9.bH9; 151

i 3.0 SURVEILLANCE REOUIREMENTS 3.5 Containment Tests (Continued) h.

Test Frecuency l

The tendon prestressing system surveillance shall be performed once every 5 years.

Basis The containment is designed for an accident pressure of 60 psig.m While the reactor is operating, the internal environment of the containment will be air at approximately atmospheric pressure and a maximum temperature of about 120*F. With these initial conditions the temperature of the steam-air mixture at the peak accident pressure of 60 psig is 288'F.

Prior to initial operation, the containment was strength-tested at 69 psig and then was leak tested. The design objective of the pre-operational leakage rate test has been established as 0.1% by weight for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 60 psig. This leakage rate is consistent with the construction of the containment, which is equipped with independent leak-testable penetrations and contains channels over all inaccessible containment liner welds, which were independently leak-tested during construction.

Safety analyses have been performed on the basis of a leakage rate of 0.1% of the free volume per day of the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the maximum hypothetical accident. With this leakage rate, a reactor power level of 1500 MWt, and with minimum containment engineered safety systems for iodine removal in operation (one air cooling and filtering unit), the public exposure would be well below 10 CFR Part 100 values in the event of the maximum hypothetical accident.* The performance of a periodic integrated leakage rate test during plant life provides a current assessment of potential leakage from the containment.

The reduced pressure (5 psig) test on the PAL is a' conservative method of testing and provides adequate indication of any potential containment leakage path. The test is conducted by pressurizing between two resilient seals on each door. The test pressure tends to unseat the resilient seals which is opposite to the accident pressure that tends to seat the resilient -

seals. The six month test ensures the overall PAL integrity at 60 psig.

The frequency of the periodic integrated leakage rate test (Type A test) is keyed to the refueling schedule for the reactor, because this test can only be performed during refueling shutdowns.

3-51 Amendment No. 68,97,139, 151 P

m

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