ML20217L052

Amend 185 to License DPR-40,revising TS to Implement 10CFR50,App J,Option B & to Allow Performance Based Changes in Conducting ILRT & Local Leak Rate Testing Types B & C & SER

ML20217L052
Person / Time
Site:	Fort Calhoun
Issue date:	03/23/1998
From:	Wharton L NRC (Affiliation Not Assigned)
To:
Shared Package
ML20217L011	List: ML20217L005 ML20217L052 ML20217L720
References
NUDOCS 9804070361
Download: ML20217L052 (19)

Text

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20045 4001

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.....p OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FOR T CALHOUN STATION. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.185 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 July 25,1997, as supplemented by letters dated November 21,1997, and "

March 3,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amendM (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: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the 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.

9804070361 980323 PDR ADOCK 05000285 P PDR

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2. Accordingly, Facility Operating License No. DPR-40 is amended by changes to the l 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 through Amendment No. 185 , are hereby incorporated in the license. The licensee i shall operate the facility in accordance with the Technical Specifications.

3. The license amendment is effective as of its date of issuance to be implemented within 30 days from the date ofissuance.

FOR THE NUCLEAR REGULATORY COMMISSION L. Raynard Wharton, Project Manager Project Directorate IV-2 Division of Reactor Projects - IlillV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 23, 1998 l

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l" ATTACHMENT TO LICENSE AMENDMENT NO. iB5 l FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 l

Revise Appendix "A" Technical Specifications as indicated below. The revised pages are  ;

identified by amendment number and contain vertical lines indicating the area of change.

l REMOVE PAGES INSERT PAGES lii lii 2-30 2-30 2-31a 2-31a 3-Ob 3-Ob 3-37 3-37 3-38 3-38 3-39 -

3-40 -

3-41 -

3-42 -

3-43 -

3-44 -

3-45 3-45 3-47 3-47 3-48 3-48 3-49 3-49 3-50 3-50 3-51 3-51 3-52 3-52 3-53a -

5-15 5-15 5-26 5-27 i

TABLE OF CONTENTS (Continued)

Pagg

,, 4.3 Nuclear Steam Supply System (NSSS) . ........... .... .. .. . ...... . ... 4-3 4.3.1 Reactor Coolant System . . . . . . . . . . ................ .... .. 4-3 4.3.2 Reactor Core and Control . . . . . . ................. ........ .4-3 4.3.3 Emergency Core Cooling . ............... .... ..... . . 4-3 4.4 Fuel Storage ................ ....... ...... . . . . . . . . . . . . 4-4 4.4.1 New Fuel Storage .. .................. ............. . .4-4 4.4.2 Spent Fuel Storage ... ........................... . . . . . . 4-4 1 4.5 Seismic Design for Class I Systems ..... ... ... ... ......... .. . .. 4-5 5.0 ADMINISTRATIVE CONTROLS . . . . . . . . . . . . . . . . . . . . . . . . . ........... 5-1 5.1 Responsibility . . . ................ ............ ......... . . 5-1 5.2 Organization .................. ............. .......... . 5-1 5.3 Facility Staff Qualifications . ..... . .............. ..... ..... . 5-la 5.4 Training . . . . . . . . . . . . . . . . . . . . . . .. . . ... .. .............. .5-3 5.5 Review and Audit ........................... ........... .. . . 5-3 5.5.1 Plant Review Committee (PRC) . . . . . . . . . . . . . . . ..... . ... . .. 5-3 5.5.2 Safety Audit and Review Committee (SARC) . . . . . . . . . . . . ....... 5-5 1

5.6 Reponable Event Action . . . . . . . . . . . . . . . . . . . . . . ....... ... . .. . 5-9 5.7 Safety Limit Violation

]

..... .................... ....... ... .5-9 '

5.8 Procedures .................. ........ . ... ............ . 5-9 5.9 Repcrting Requirements ............................ .. ....... 5-10  ;

5.9.1 Routine Reports . . . . . . . . . . . . . . . . . . . . . . . . . . .......... . 5-10 5.9.2 Reportable Events . . ..... ........................ ... . 5-12 5.9.3 Special Reports .................................... .. 5-15 5.9.4 Unique Reporting Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-15 5.9.5 Core Operating Limits Report .. .......... ........ ... .. . 5-17a 5.10 Records Retention . . . . . . . . . . . . . . . . . ....... .............. . . 5-18 5.11 Radiation Protection Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 19 5.12 DELETED 1 5.13 Secondary Water Chemistry . .......... ...... .. .. .......... . 5-20 5.14 Systems integrity . . . . . . . . . . . . . . . . . . . . ..... ............. .5-21 5.15 ' Post-Accident Radiological Sampling and Monitoring ........ .. . ..... .. .. 5-21 5.16 Radiological Effluents and Environmental Monitoring Programs . . . . . . . ..... . 5-22 5.16.1 Radioactive Effluent Controls Program . . . . . . . . . . ... ......... . 5-22 5.16.2 Radiological Environmental Monitoring Program . . . ...... .. . .... 5-23 l

5.17 Offsite Dose Calculation Manual (ODCM) . .... ...... ..... .. . . . . . 5-24 5.18 Process Control Program (PCP) .... ...................... ...... 5-25 l 5.19 Containment Leakage Rate Testing Program . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-26 l l 6.0 INTERIM SPECIAL TECHNICAL SPECIFICATIONS . . . . . . . . . . . . . . . . . . . . . . . . 6-1 l

l 6.1 DELETED 6.2 DELETED 6.3 DELETED 6.4 DELETED iii Amendment No. 32,34,.54,55,57,

2. 3.on_ ,a_t_ , n o. .o. _s ,n. o. , u i . i. _c,

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2.0 : LIMITING CONDITIONS FOR OPERATION

'- Conainment System 2.6 -

Aeolicability Applies to the reactor containment system.

Obiective To assure the integrity of the reactor containment system.

Soecifications (1) Containment Integrity -

a. Containment integrity shall not be violated unless the reactor is in a cold or refueling shutdown condition. Without containment integrityi 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 subcritical 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 j opened intermittently under administrative control without constituting a ,

violation of containment integrity. I

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 ased 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 5.19. I (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. Otherwise, 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 met.

(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 />.

2-30 Amendment No. 68,138,151, 185 i

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2.0 LIMITING CONDITIONS FOR OPERATION 2.6 Containment System (Continued)

.. Basit 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. The shutdown margins are selected based on the type of l

activities that are being carried out. The refueling boron concentration provides a shutdown margin which precludes criticality under any circumstances. Each CEDM must I be tested and some have two CEA's attached.

t 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.* 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 l 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 l 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 5.19. l Specification 2.6(1)b(ii) applies when mechanisms other than a door, such as the inner door equalizing valve, are declared inoperable.

l The Hydrogen Purge System is required to be operable in order to control the quantity l

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 l the system is capable of performing its design function. The blowers (VA-80A and VA-80B), associated valves, and piping are single failure proof, have been designed as a Seismic Class I System, and are redundant to the VA-82 filter header. VA-80A or VA-l 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 j 45 flammability limit.* Electrical Equipment qualification was not required as the L radiation doses in the area'of the Hydrogen Purge System equipment were below the minimum requirements.*

l- 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 reinoved from operation while the other train has the capability to provide 100% hydrogen control.

References (1) USAR, Section 14.16; Figure 14.16-2 (2) Regulatory Guide 1.7 (1971)

(3) USAR, Section 14.17 l

l (4) Engineering Study 86-10, Calculation 53 i l 2-31a Amendment No. 135,151, 185 l l

T 3.0 SURVEILLANCE REOUIREMENTS BASIS l .-

Specifications 3.0.1 through 3.0.4 establish the general requirements applicable to Surveillance Requirements. These requirements are based on the Surveillance Requirements stated in the Code of Federal Regulations,10 CFR 50.36(c)(3):

l " Surveillance requirements are requirements relating to test, calibration, or

-inspection to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting condition of operation will be met."

Specification 3.0.1 establishes the limit for which the specified time interval for Surveillance Requirements may be extended. It permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g.,

[. transient conditions or other ongoing surveillance or maintenance activities. It also provides flexibility to accommodate the length of a fuel cycle for surveillances that are performed at each refueling outage and are specified with an 18-month surveillance interval. It is not intended that this provision be used repeatedly as a convenience to i-extend surveillance intervals beyond that specified for surveillance that are not performed during refueling outages. The limitation of Specification 3.0.1 is based on engineering l- judgement and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements.

l This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.

The provisions of Specification 3.0.2 defm' e the surveillance intervals for use in the l

Technical Specifications. This clarification is provided to ensure consistency in surveillance intervals.throughout the Technical Specifications. A few surveillance

. requirements have uncommon intervals. In such a case the surveillance interval shall be performed as defined by the individual specifications.

Specification 3.0.3 extends the testing interval required by codes and standards l referenced by the Technical Specifications. This clarification is provided to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities. Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the codes and standards referenced therein. The requirements of regulations take precedence over the TS. Therefore test intervals governed by regulation cannot be extended by the TS. An example of this exception is the Containment Leakage Rate Testing Program.

l Specification 3.0.4 establishes the failure to perform a Surveillance Requirement within the allowed surveillance interval, as defined by the provisions of Specifications 3.0.1 and 3.0.2, as a condition that constitutes a failure to meet the OPERABILITY requirements for the corresponding Limiting Condition for Operation. Under the provisions

. 1 3-Ob Amendment No. 122,129,,152,185 l t

~ 3.0 - SURVEILLANCE REOUIREMENTS 3.5 - Containment Test Aeolicability

. Applies to containment leakage and structural integrity.

Obiective

' To verify that the:

(1) Locked closed manual containment isolation valves are closed and locked, (2) potential leakage from containment is within acceptable limits, and (3) structural performance of all .important components 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 Integrated Leakane Rate Test (Type A Tests)-

Perform required visual examinations and leakage rate testing in accordance with -

the Containment Leakage Rate Testing Program.

(3) .' Containment Penetrations Leak Rate Tests (Ti B Tests)

Perform required visual examinations and leakage rate testing in accordance with the Containment Leakage Rate Testing Program for the following penetrations:

'(i)- Equipment Hatch (ii) Personnel Access Lock (iii) Mechanical Penetrations M-1 through M-99 (iv) Fuel Transfer Tube (Mechanical Penetration M-100)

_v)

( Electrical Penetrations:

l A-1 ' B-9 D-6 F-2 E-HCV-383-3A A-2 .B-10 D-7 F-4 E-HCV-383-3B A-4 B-11 D-8 F-5 FeHCV-383-4A A-5 C-1 D-9 F-6 E-HCV-3834B A-6 C-2 D-10 F-7 A-7 C-4 D-11 F-8 A-8 C-5 E-1 F-9

' A-9 C-6 E-2 F-10 3-37 Amendment No. 9&d54,185 l- 1

i 3.0 SURVEILLANCE REOUIREMENTS 3.5 Containment Testi (Continued)

A-10 C-7 E-4 F-11 A-11 C-8 E-5 G-1 B-1 C-9 E-6 G-2 B-2 C-10 E-7 G-3 B-4 C-11 E-8 G-4 i B-5 D-1 E-9 H-1 B-6 D-2 E-10 H-2 B-7 D-4 E-11 H-3 B-8 D-5 F-1 H-4 (4) Containment Isolation Valves Leak Rate Tests (Type C Tests)

Perform required visual examinations and leakage rate testing in accordance with the Containment Leakage Rate Testing Program for the following penetrations:

M-2 M-31 M-52 IA-3092 M-7 M-38 M-53 IA-3093 l

M-8 M-39 M-57 IA-3094 M-11 M-40 M-58 M-14 M-42 M-69 l 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 M-24 ' M-48 M-88 M-25 M-50 M-HCV-383-3 M-30 M-51 M-HCV-383-4 l

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3-38 Amendment No. 95dM,185

, (Next Page is 3-45) l.

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3.0 SURVEILLANCE REOUIREMENTS 3.5 Containment Tests (Continued)

(5) Surveillance for Prestressing System l

( a. Samole Selection l

The 210 dome tendons and 616 helical wall tendons shall be periodically inspe::ted 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 observed data.

b. Visual Insnection 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 leakage, and other significant structural deterioration or disintegration.

(ii) For each surveillance tendon, selected in accordance with 3.5(5)a., the l tendon anchorage assembly hardware shall be visually inspected for signs of abnormal material behavior or wear.

(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 l inspected to detect grease leakage or grease cap deformations. Removal of the grease caps is not necessary for this inspection.

c. Prestress Monitoring Tests i Liftoff tests shall be performed on each tendon selected in accordance with

. 3.5(5)a. to monitor prestress. Additionally, the tests shall include the following: l 3-45 Amendment No. 95,97,139,151,185

3.0 SURVEILLANCE REOUIREMENTS 3.5 Containment Tests (Continued)

The tendons detensioned in accordance with 3.5(5)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 broken shall be removed for tensile testing and visual examination. ,

e. Insoection 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-S2 D.- " Methylene Blue Method," Standard Methods for Examination of Water 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. Acceotance Criteria (i) No evidence of significant structural deterion m of the concrete inspected in accordance with 3.5(5)b.(i) and 3.5(5)b. >ii) which may affect l the structural integrity of the containment structure c.n be detected.

i l 3-47 Amendment No. 95,97,139,151,185 l

3.0 SURVEILLANCE REOUIREMENTS 3.5 Coneninment 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. i 1

l 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(5)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 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.

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3-48 Amendment No. 95,97,139,151,185

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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 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(5)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(5)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(5)g. I 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 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(5)g. l (iii) If during the detensioning and retensioning of tendons in accordance with 3.5(5)c., the elongation corresponding to a specific load differs by more l 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(5)g. l 3-49 Amendment No. 95,97,139,151,185

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 for Type BA wire. Failure in the tensile test at strength or elongation values less than those specified shall be considered reportable per 3.5(5)g. l Other conditions which indicate corrosion found by visual examination of the wire shall be considered reportable per 3.5(5)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) Chloiides i 10 ppm (c) Nitrates A 10 ppm (d) Sulfides i 10 ppm I

(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(5)g. l

g. Corrective Action and Reporting If the above acceptance criteria are not met, an immediate investigation shall be made to determine the cause(s) and extent of the non-conformance 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.

l 3-50 Amendment No. 24,68,95,139,151185

p 3.0 SURVEILLANCE REOUIREMENTS -

3.5 Containment Tests (Continued) -

l ..

h. Test Freauency

! The tendon prestressing system surveillance shall be performed once every l- 5 years. l l Basis The containment is designed for an accident pressure .of 60 psig.* 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 i

steam-air mixture at the peak accident pressure of 60 psig is 288 F.

t 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 an integrated leakage rate test and performance of local leak rate testing of individual penetrations at periodic intervals 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. A periodic test ensures the overall PAL integrity at 60 psig. l l The integrated leakage rate test (Type A test) can only be performed l during refueling shutdowns. I l

l 3-51 Amendment No. 58,97,139,151,185

3.0 SURVEILLANCE REOUIREMENTS 3.5 Containment Tests (continued)

The frequency of periodic integrated leakage rate tests is based on several major considerations: (1) There is a low probability of leaks in the liner ,

because of the test of leak-tightness of the welds during erection and conformance of the complete containment to a low leak rate at 60 psig during pre-operational testing, which is consistent with 0.1% leakage at design basis accident conditions and absence of any significant stresses in the liner during reactor operation. (2) Periodic testing is conducted at full accident pressure, on those portions of the containment envelope that are i most likely to develop leaks during reactor operation (penetrations and l isolation valves). A low value (0.60 L ) of total leakage is specified as acceptable from penetrations and isolation valves. (3) The tendon stress surveillance program provides assurance that an important part of the structural integrity of the containment is maintained. (4) A review of leakage rates obtained during past containment integrated leakage rate testing is conducted to set appropriate frequency of performance not to exceed once every 10 years. (5) Visual inspection of the containment structure is conducted every other refueling and prior to each Integrated Leakage Rate Test.

As left leakage prior to the first startup after performing a required f l leakage test is required to be < 0.6 L, for combined Type B and C leakage, and < 0.75 L, for overall Type A leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 L,. At s 1.0 L, the offsite dose consequences are bounded by the assumptions of the safety analysis.

Integrity tests of the purge isolation valves are established to identify l excessive degradation of the resilient seats of these valves. Simultaneous testing of redundant purge valves from a leak test connection accessible from outside containment provides adequate testing. The tet, ting method is identical to the Type C purge isolation valve test performed in accordance with 10 CFR Part 50, Appendix J. For leakages found to be greater than 18,000 SCCM, repairs shall be initiated to ensure these valves meet the acceptance criteria.

A reduction in prestressing force and changes in physical conditions are expected for the pmstressing system. Allowances have been made in the  !

reactor ' building design for the reduction and changes. Through i comparisons between the documented inspection results and the initial quality control records, the reductions in prestress and the physical changes are trended to verify excessive reductions or changes do not occur or are detected in a timely manner to be corrected.

l 3-52 Amendment No. 9hn9,185

5.0 ADMINISTRATIVE CONTROLS 5.9.3' Special Recons Special reports shall be submitted to the Regional Administrator of the appropriate NRC Regional Office within the time period specified for each repon. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification where appropriate:

a. In-service inspection report, reference 3.3.

b. Tendon surveillance, reference 3.5.

c. Containment structural tests, reference 3.5.

d. Special maintenance reports.

e. DELETED g

f. DELETED

g. Materials radiation surveillance specimens repons, reference 3.3.

h. DELETED

i. Post-accident monitoring instrumentation, reference 2.21

j. Electrical systems, reference 2.7(2).

5.9.4 Uniaue Reportina Reauirements

a. Annual Radioactive Effluent Release Report The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year of operation shall be submitted before May 1 of each year. The report shall include a summary of the quantities of  ;

radioactive liquid and gaseous effluents and solid waste released from the unit.

The material provided shall be 1) consistent with the objectives outlined in the ODCM and PCP, and 2) in conformance with 10 CFR 50.36a. and Section IV.B.1 of Appendix I to 10 CFR 50.

b. Annual Radiological Environmental Operation Report The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2)Section IV.B.2, IV B.3, and IV.C of Appendix I to 10 CFR 50.

c. Fire Protection Program Deficiency Report Deficiencies in the Fire Protection Program described in the Updated Safety Analysis Report which meet the reportability criteria of 10 CFR 50.73 shall be reported pursuant to Section 5.9.2 of the Technical Specifications.

5-15 Amendment No. 9,24,38,46,86,110, (Next page is 5-17a) 113,133,147,152,160,164, 185 i

i I

5.0 ADMINISTRATIVE CONTROLS 5.19 Containment Leakane Rate Testine Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J. Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program, dated September 1995," as modified by the following exceptions:

(1) If the Personnel Air Lock (PAL) is opened during periods when containment integrity is not required, the PAL door seals shall be tested at the end of such periods and the entire PAL shall be tested within 14 days after RCS temperature Ta > 210*F.

(2) Type A tests may be deferred for penetrations of the steel pressure retaining boundary where the nominal diameter does not exceed one inch.

(3) Elapsed time between consecutive Type A tests used to determine performance shall be at least 24 months or refueling interval.

The containment design accident pressure (P,) is 60 psig.

The maximum allowable primary containment leakage rate, L , at P , shall be 0.1% of containment air weight per day.

Leakage Rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is s 1.0 L,. During unit startup following testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 L, Maximum Pathway Leakage Rate (MXPLR) for Type B and C tests and s 0.75 L, for Type A tests,

b. Personnel Air Lock testing acceptance criteria are:

(1) Overall Personnel Air Lock leakage is s 0.1 L, when tested at 'n P,.

(2) For each PAL door, seal leakage rate is s 0.01 L, when pressurized to 2 5.0 psig.

c. Containment Purge Valve (PCV-742A/B/C/D) testing acceptance criterion is:

For each Containment Purge Valve, leakage rate is < 18.000 SCCM when tested at 2 P,.

d. If at any time when containment integrity is required and the total Type B and C measured leakage rate exceeds 0.60 L, Minimum Pathway Leakage Rate (MNLPR), repairs shall be initiated immediately. If repairs and retesting fail to demonstrate conformance to this acceptance criteria within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, then containment shall be declared inoperable.

5-26 Amendment No.185 l

l i

l l

c l

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5.0 ADMINISTRATIVE CONTROLS

! 5.19 Containment Leakane Rate Testing Program (continued)

The provisions of Specification 3.0.1 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of Specification 3.0.4 are applicable to the Containment Leakage Rate Testing Program.

l 5-27 Amendment No.185

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