ML20247K698

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Amends 23 & 4 to Licenses NPF-68 & NPF-81,respectively, Revising Tech Specs Re Containment Tendon Surveillance
ML20247K698
Person / Time
Site: Vogtle  Southern Nuclear icon.png
Issue date: 09/12/1989
From: Matthews D
Office of Nuclear Reactor Regulation
To:
Georgia Power Co, Oglethorpe Power Corp, Municipal Electric Authority of Georgia, City of Dalton, GA
Shared Package
ML20247K704 List:
References
NPF-68-A-023, NPF-81-A-004 NUDOCS 8909210245
Download: ML20247K698 (13)
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UNITED STATES

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NUCLEAR REGULATORY COMMISSION

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GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTIC AUTHORITY OF' GEORGIA CITY OF DALTON, GEORGIA V0GTLE ELECTRIC GENERATING PLANT, UNIT 1

. AMENDMENT TO FACILITY OPERATING LICENSE Amenoment No. 23' License No. NPF-68

1. -

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

A.

The application for amendment to the Vogtle Electric Generating Plant, Unit 1 (the facility) Facility Operating License No. NPF-68 filed by the Georgia Power Company acting for itself, Oglethorpe Power Corpo-ration, Municipal Electric Authority of Georgia, and City of Dalton, Georgia, (the licensees) dated May 9,1989 as supplemented July 28 and August 14, 1989, complies with the. standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules an8 regulations set forth in 10 CFR Chapter I;-

5.

The facility will operate in ccnformity with the application, as amended, 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 set forth in 10 CFR Chapter I; D.

The issuance of this license amendment will not be inimical to the commo'n defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

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Accordingly, the license ic hereby amended by page changes to the Technical i

Specifications as indicated in the attachments to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-68 is hereby amended to read as follows:

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(2)- Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 23, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. GPC shall operate the facility in acccrdance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Original signed by:

David B. Matthews, Director Project Directorate 11-3 i

Division of Reactor Projects I/II Office of Nuclear Reactor Regulation 1

Attachment:

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Technical Specification Changes Date of Issuance:

September 12, 1989 I

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0FFICIAL RECORD COPY I

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GEORGIA POWER COMPANY CGLETHORPE POWER CORPORATION MUNICIPAL ELECTIC AUTHORITY OF' GEORGIA CITY OF DALTON, GEORGIA

- V0GTLE ELECTRIC GENERATING PLANT, UNIT 2

- AMENDMENT TO FACILITY OPERATING LICENSE 1

Amendment No. 4 License No. NPF-81 1.

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

The app (lication for amendment to the'Vogtle Electric Generating Plant, A.

Unit 2 the facility) Facility Operating License No. NPF-81 filed by the Georgia Power Company acting for itself, Oglethorpe Power Corpo-ration, Municipal Electric Authority of Georgia, and City of Dalton, Georgia, (the licensees) dated May 9,1989 as supplemented July 28 and August 14,-1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules ano regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized by-this amendment can be conducted without endangering the bealth and -

safety of the public, and (ii) that such activities will be cofiducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; 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; t-and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

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Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachments to this license amendment and paragraph 2.C.(2) of Facility Operating License No.. NPF-81 is hereby amended to read as follows:

(2) ~ Technical Specifications and Environmental Protection Plan The Technical Specifications; contained in Appendix A, as revised -

through Amendment No. 4, and.the Environmental Protection Plan contained in. Appendix B, both of which are attached. hereto, are hereby. incorporated into.this-license. GPC shall operate'the i

facility in accordance with:the Technical Specifications and the l

Environmental Protection Plan.

3.

This license anendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

FOR.THE NUCLEAR REGULATORY COMMISSION Original signed by:

David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation.

Attachment:

Technical Specification Changes Date of Issuance: September 12, 1989 i

OFFICIAL RECORD COPY

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,4 ATTACHMENT TO LICENSE ' AMENDMENT NO. 23 FACILITY OPERATING LICENSE NO. NPF-68 AND LICENSE AMENDMENT NO. -4 '

FACILITY OPERATING LICENSE NO. NPF-81 DOCKET NOS. 50-424 AND 50-425' Replace.the following pages of the Appendix "A" Technical Specifications with the enclosed pages..The revised pages are identified by Amendment number and.

contain vertical lines indicating the area of change. The corresponding overleaf page is also provided to maintain document completeness.

Amended Page Overleaf Page

~3/4 6-8 3/46-7 3/4 6-9 3/4 6-9a 3/4 6-9b' 3/4 6-10 8 3/4 6-2 B 3/4 6-1

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$6 CONTAINMENT SYSTEMS.

AIR TEMPERATURE' LIMITING CONDITION FOR OPERAIION

. 3. 6; 1. 5 -: Primary containment average air temperature (TE-2S63, TE-2612,.

TE-2613).shall not exceed 120*F.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

- With the containment average air temperature greater than 120*F, reduce the average air-temperature to within the limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 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 />.

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SURVEILLANCE REQUIREMENTS 4.6.1.5 The primary containment average air temperature shall be the arith-metical average of the temperatures at the following locations and shall be determined at least once per 2' hours:

Location Tag Numbers

  • a.

Level 2 TE-2563 b.

Level B TE-2613 c.

Level C TE-2612

  • 0r local sample at corresponding location V0GTLE UNITS - 1 & 2 3/4 6-7

CONTAINMENT SYSTEMS CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containments shall be maintained at a f

level consistent with the acceptance criteria in Specification 4.6.1.6.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

a.

With the abnormal degradation indicated by the conditions in Specification 4.6.1.6.la.4, restore the containment (s) to the required level of integrity or verify that containment integrity is maintained within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and perform an engineering evaluation of the containment (s) and provide a Special Report to te e Commission within 15 days in accordance with Specification 6.9.2 or be in at least HDT 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 />.

b.

With the indicated abnormal degradation of the structural integrity other than ACTION a. at a level below the acceptance criteria of Specification 4.6.1.6, restore the containment (s) to the required level of integrity or verify that containment integrity is maintained within 15 days; perform an engineering evaluation of the containment (s) and provide a Special Report to the Commission within 30 days in accordance with Specification 6.9.2 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 />.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIF.EMENTS 4.6.1.6 CONTAINMENT PRESTRESSING SYSTEM The structural integrity of the prestressing tendons of the containments shall be demonstrated at the end of 1, 3, and 5 years following the initial containment vessel structural integrity test and at 5 year intervals thereafter.

For combined inspections of two containments in a plant, lift-off l

testing will be performed in accordance with the inspection schedule shown in Figure 3.6-1.

4.6.1.6.1 The adequacy of prestressing forces in tendons shall be demonstrated by:

l a.

Determining that a random but representative sample of at least 11 tendons (7 hoop and 4 inverted-U) each have an observed lift-off V0GTLE UNITS - 1 & 2 3/4 6-8 Amendment No.23 (Unit 1)

Amendment No. 4 (Unit 2) u _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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.c, CONTAINMENT SYSTEMS

-SURVEILLANCE REQUIREMENTS (Continued) force within predicted limits established for each tendon.

For each subsequent inspection, one tendon.from each group.shall.be kept unchanged to develop a history and to correlate the observed data.

The procedure of inspection and the tendon acceptance criteria shall be as follows:

(1)' If the measured prestressing force of the selected tendon in,a l

group-lies above the prescribed lower limit, the lift-off test L

is considered to be a positive indication of the' sample tendon's acceptability.

(2)

If the measured prestressing force of the: selected tendon in a..

i group lies between the prescribed lower limit and 90% of the prescribed lower limit, two tendons, one'on each side of this tendon, shall be checked for their'prestressing forces.

If the prestressing forces of these two tendons'are above 95% of the prescribed lower limits for the tendons, all three tendons shall be restored to the required level of integrity,' and the.-

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I' tendon group shall be considered as acceptable.

If the measured prestressing force of any two tendons falls below 95%

of the prescribed lower limits of the tendons, additional lift off testing shall be done to detect the cause and extent of such occurrence.

The conditions shall be considered as an indication of abnormal degradation of the containment structure.

(3) If the measured prestressing force of any tendon lies below 90%

of the prescribed lower limit, an engineering investigation.

will be performed to determine the cause and extent of the occurrence.

The condition shall be considerect as an indication of abnormal degradation of the containment structure.

(4) If the average of all measured prestressing forces for each group (corrected for average condition) is found to be less than the minimum required prestress level at anchorage location for that group, the condition shall be considered as abnormal degradation of the containment structure.

(5) If from consecutive surveillance the measured prestressing forces for the same tendon or tendons in a group indicate a trend of prestress loss larger than expected and the resulting prestressing forces will be less than the minimum required for the group before the next scheduled surveillance, additional lift off testing shall be done so as to determine the cause and extent of such occurrence.

The condition shall be considered as an indication of abnormal degradation of the containment structure.

(6) Unless there is abnormal degradation of the containment ves>el during the first three inspections, the sample population for subsequent inspections shall include at least 7 tendons (4 hoop, and 3 inverted-U).

b.

Performing tendon detensioning, inspections, and material tests on a previously stressed tendon.

Two tendons, one from each group, shall V0GTLE UNITS - 1 & 2 3/4 6-9 Amendment No.23 (Unit 1)

Amendment No. 4 (Unit 2)

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ll-CONTAINMENT SYSTEMS 1

SURVEILLANCE REQUIREMENTS (Continued) be detensioned on Unit 1 each time lift-offs are performed on Unit 1 per Figure 3.6-1.

One tendon shall be detensioned on Unit 1 each l

time lift offs are performed on Unit 2 per Figure 3.6-1.

A randomly l

selected tendon shall be essentially completely detensioned in order to identify broken or damaged wires and determining that over the entire length of the removed wire sample (which should include the broken wire if so identified) that:

l (1) The tendon wires are free of corrosion, cracks, and damage, and (2) A minimum tensile strength of 270,000 psi (guaranteed ultimate strength of the tendon material) exists for at least three wire samples (one from each end and one at mid-length) cut from each removed wire.

Failure to meet the requirements of 4.6.1.6.lb shall be considered as an indication of abnormal degradation of the containment structure.

c.

For Unit 1 only, performing tendon retensioning of detensioned tendons as close as possible to their observed or predicted lift-off force, whichever is greater but not to exceed a stress level of 70%

of the guaranteed ultimate tensile strength (GUT 5) for the tendon material.

During retensioning of thqse tendons, the changes in load and elongation should be measured simultaneously at a minimum of three approximately equally-spaced levels of force between zero and the seating force.

If the elongation corresponding to a specific load differs by more than 10% from that recorded during the installation, an investigation should be made to ensure that the difference is not related to wire failures or slip of wires in anchorages.

This condition shall be considered as an indication of abnormal degradation of the containment structure.

d.

Verifying the OPERABILITY of the sheathing filler grease by assuring:

(1) There are no changes in the presence or physical appearance of the sheathing filler grease including the presence of free water.

(2) Amount of grease replaced does not exceed 5% of the net duct volume, when injected at a pressure not to exceed the designer's specifications.

(3)

Minimum grease coverage exists for the different parts of the anchorage system.

(4) During general visual examination of the containment exterior surface, grease leakage that could affect containment integrity is not present, and V0GTLE UNITS - 1 & 2 3/4 6-9a Amendment No. 23 (Unit 1)

Amendment No. 4 (Unit 2)

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CONTAINMENT-SYSTEMS SURVEILLANCE' REQ'UIREMLNTS'(Continued) b (6) -The' chemical properties of'the filler material.are within the tolerance limits specified as follows:

Water Content 0-10% (by dry wt.)

Chlorides 0-10 ppm Nitrates 0-10 ppm.

Sulfides 0-10 ppm Reserved Alkalinity

>50% of.the installed value;'

(Base Numbers)

Failure to meetirequirement'of 4.6.1.6.1d shall be considered as an indication of abnormal degradation of the containment structure.

4.6.1.6.2: E.vf inchorages and Adjacent Surfaces.

The structural integrity of the end anchorages of all tendons inspected pursuant to Specification 4.6.1.6.1 and tiec adjacent surfaces shall be. demonstrated by determining through visual-inspectim that no apparent changes have occurred.

a All end an:f.orages including anchor blocks, wedges, shims, and bearing p'atss:

inspect for moisture, corrosion and cracks, and for-warping of bearing plates.

b.

Concrete surfaces adjacent to hoop tendon anchorages: ' inspect for moisture, corrosion, distortion, and cracking.

c.

Steel plating surrounding the inverted-U tendon anchorages:

inspect for moisture, corrosion, distortion, and cracking.

Significant grease leakage,' grease cap deformation, or abnormal concrete / steel plating conditions shall be considered as an indication of abnormal degradation of containment structure.

4.6.1.6.3 Containment Surfaces. The exterior surface of the containments should be visually examined to detect areas of large spall, severe scaling, D-cracking, other surface deterioration or disintegration, or significant grease leakage, each of which can be considered as evidence of abnormal degradation of structural integrity of the (. containments.

P V0GTLE UNITS - 1 & 2 3/4 6-9b Amendment No. 23 (Unit 1)

Amendment No. 4 (Unit 2)

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25 3,5 TIME AFTER INITIAL STRUCTURAL INTEGRITY TESTING OF CONTAINMENT, YEARS (Litt-off Testing Schedule, Containment No.1) g 3

y 1,5 2,5 3,5 TIME AFTER INITIAL STRUCTURAL INTEGRITY TESTING OF CONTAINMENT, YEARS (Lift-off Testing Schedule, Containment No. 2)

Schedule to be used provided:

a. The containments are identical in'all aspects such as size, tendon system, design, materials of construction, and method of construction.

The tendon system for Unit 2 does not provide for detensioning. Detensioning can be performed only on the Unit 1 tendon system.

b. The 1-year inspection for Unit 2 will consist of a visualinspection only. No lift-off testing will be performed on Unit 2 until the 3-year inspection.
c. There is no unique situation that may subject either containment to a different potential for structural or tendon deterioration.
d. The Unit 1 and Unit 2 surveillance may be performed back to-back to facilitate detensioning of Unit 1 tendons during the Unit 2 surveillance.

e in order to perform back-to-back surveillance on Units 1 and 2, the Unit 1 10-year surveillance and the Unit 2 5-year surveillance are to be performed between 5/1/95 and 11/1/95.

FIGURE 3.6-1 SCHEJULE OF LIST-0FF TESTING FOR TWO CONTAINMENTS AT A SITE V0GTLE UNITS - 1 & 2 3/4 6-10 Amendment No.23 (Unit 1)

Amendment No. 4 (Unit 2)

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive.

materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses.

This restriction, in conjunction with the leakage rate limitation, will limit the SITE BOUNDARY radiation doses to within the dose guideline values of 13 CFR Part 100 during accident conditions.

3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, P,.

As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L, during performance of the periodic test to account for possibledegradationofthecontainmentleakagebarriersbitweenleakagetests.

The surveillance testing for measuring leakage rates is consistent witn the requirements of Appendix J of 10 CFR Part 50.

3/4.6.1.3 CONTAINMENT AIR LOCKS 1

The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate.

Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.

l 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that:

(1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 3 psig, and (2) the containment peak pressure does not exceed the design pressure of 52 psig during steam line break conditions.

The maximum peak pressure expected to be obtained from a steam line break event is 41.9 psig assuming an initial containment pressure of 0.3 psig.

The initial positive containment pressure will limit the total pressure to less than P, which is less than design pressure and is consistent with the safety 3

analyses.

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I V0GTLE UNITS - 1 & 2 B 3/4 6-1

A CONTAINMENT SYSTEMS BASES 3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature does not exceed the initial temperature condition assumed in the safety analysis for a steam line break accident. Measurements shall be made at all listed locations, whether by fixed or portable instruments, prior to determining the average air temperature.

3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility.

Structural. integrity is required to ensure that the containment will withstand the maximum pressure of 41.9 psig in the event of a steam line 4

break accident.

The measurement of containment tendon lift-off force, the ten-sile tests of the tendon stra.vjs for Unit 1, the visual examination of tendons, anchorages and exposed interior and exterior surfaces of the containment and the Type A leakage test for both units are sufficient to demonstrate this capa-I bility.

(The tendon strand samples will also be subjected to stress cycling

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tests and to accelerated corrosion tests to simulate the tendon's operating i

conditions and environment.) Lift-off testing on Unit 2 will be accompanied by detensioning of one tendon on Unit 1.

This tendon will alternate between the hoop and inverted -0 tendons.

With regard to D-cracking, the acceptance criteria for the visual inspection of the containment concrete is that the area comprising 0-cracking should not exceed 25 sq. ft.

The conditions referenced by Action statement 3.6.1.6.b do not define I

abnormal containment degradation.

These conditions are indications of potential abnormal degradation and their existence requires an appropriate engineering evaluation and a Special Report in accordance with Specification 6.9.2.

The required Special Reports from any engineering evaluation of contain-J ment abnormalities shall include a description of the tendon condition,' the l

condition of the concrete (especially at tendon anchorages), the inspection j

procedures, the tolerances ~~on cracking, the results of the engineering evalua-tion, and the corrective actions taken, or proposed.

l 3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM The 24-inch containment purge supply and exhaust isolation valves are required to be sealed closed during plant operations since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident.

Maintaining these valves sealed closed during plant operation ensures that exces-sive quantities of radioactive materials will not be released via the Containment I

Purge System.

To provide assurance that these containment valves cannot be inad-I vertently opened, the valves are sealed closed in accordance with Standard Review Plan 6.2.4.

Sealed closed isolation valves are isolation valves under admini-strative control to assure that they cannot be inadvertently opened.

Admini-strative control includes mechanical devices to seal or lock the valve closed, l

the use of blind flanges, or removal of power to the valve operator.

l V0GTLE UNITS - 1 & 2 8 3/4 6-2 Amendment No. 23(Unit 1)

Amendment No. 4(Unit 2)

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