ML19309B908

Tech Spec Change Request 77A to App a of DPR-50 Re Revision of Tendon Surveillance to Reflect Reg Guide 1.35,Revision 3 Requirements & to Delete Ring Girder Surveillance Program

ML19309B908
Person / Time
Site:	Crane
Issue date:	03/31/1980
From:	METROPOLITAN EDISON CO.
To:
Shared Package
ML19309B905	List: ML19309B904 ML19309B906 ML19309B908
References
NUDOCS 8004070325
Download: ML19309B908 (10)

Text

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O UNITED STATES OF AMERICA NUCLEAR PEGULATORY COMMISSION IN THE MATTER OF DOCKET No.

50-289 LI'P.NSE NO. DPR-50 METROPOLITAN EDISON COMPANY This is to certify that a copy of Technical Specification Change Request No. 77A to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with the U. S. Nuclear Regulatory Commission and been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania and Dauphin County, Pennsylvania by deposit in the United States mail, addressed as follows:

Mr. Weldon B. Arehart Mr. Harry B. Reese, Jr.

Board of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Court House Middletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 METROPOLITAN EDISON COMPANY 5

Y I Vice President Dated:

March 31, 1980 8004070 3 2 5

iThi 22 MilejIsland Huclear Station,: Unit 'l-10perating; License No.'DPR-50

<Dockst No.!50-289' f'. ~ Techinical Spbhification Change Request No. 77A

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- The licensee requests *that-the attached changed pages replace Jpages 11, y,

vii',

h-35 through.'h-36f, figuresch~.k-l~through h.h-5, tables k.h-1:through h.h-3,

' ;and page 6-18.~ ;This Amendment. Change. Request supercedes T.S.C.R.-No. 77

-Reason'for Change-Request

- This1 Amended Change Requestkis being ' submitted to revise the tendon surveillance -

" specification to. reflect' Regulatory. Guide 1.35 Proposed Rev. 3, and to delete

<the Ring Girder' Surveillance' Program.

Safety: Evaluation Justifying the Change This Change. Request has been preparedito reflect -the surveillance requirements of Proposed Revision 3 of NRC Regulatory Guide 1 3h, and the ACI-ASNS Article CC-9000 Section;CC-9222.2. This change does not constitute and unreviewed safety question in that it only modifies the tendon surveillance program.

i Additionally, as; stated in our submittal of Technical Specification Change Request, the ~ Commission ste ted -that, as a result of their reviev 'of the previous Reactor -

Containment Building. Ring Girder Surveillance Tests, the safety margins in the

. structure at the'present time, are adequate.

~The Commission has concluded that there is reasonable assurance that the containment structure vill continue,to perform its intended safety function.

(In their letter of December.29, 1977, the Commission requested-to be informed of the 1 repairs committed to in our. letter of June 3, 1977 These repairs have been completed in strict accordance-vith the repair procedure for.the cleaning, packing and curing of exposed concrete voids adjacent.to dome bearing plates.

License Amendment Fee (10 CFR 170.22)

-This is an amendment to a previously submitted change request, therefore, no fee is required.

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! TABLE'0F CONTENTS 4 y

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,~:Section(

P_ age 7$Y ' <

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LMINIMUM CONDITIONS FOR CRITICALITY' Ac^

31 hi*

fREACTOR COOLANT SYSTEM ACTIVITY 3 3.1 5

- ; CHEMISTRY 3-10 w.

-3 1.63 i LEAKAGE..

3-16 _-

c:

< 3.1. 7.

? MODERATOR: TEMPERATURE COEFFICIENT OFfREACTIVITY.

3-12

' 3.1. 8 :

SINGLE LOOPJRESTRICTIONS.

3-17' 3.1 9' LIDW POWER PHYSICS TESTING RESTRICTIONS 18-1

3.1;10 1 CONTROL ROD OPERATION 3-18a 3.1.11

. REACTOR INTERNALS: VENT VALVES

'3 18b

~ 3. 2 MAKEUP AND PURIFICATION'AND CHD4ICAL ADDITION' SYSTEMS 3-19

~

3.3; 1 EMERGENCY CORE COOLING, REACTOR BUILDING EMERGENCY COOLING, AND REACTOR-BUILDING SPRAY SYSTEMS

.3 ;

.3.k.

' TURBINE CYCLE 3 35 INSTRUMENTATION SYSTEMS-3-27

-3 5 1" OPERATIONAL SAFETY INSTRUMENTATION:

3-27

'352 CONTROL ROD GROUP AND. POWER DISTRIBUTION LIMITS 33-3'5 3.

ENGINEERED SAFEGUARDS PROTECTION SYSTEM ACTUATION-.

SETPOINTS 3-37 3.5.h

~ INCORE INSTRUMENTATION 3 1 3.6 REACTOR BUILDING 3-h1 3.7

-UNIT ELECTRICAL POWER SYSTEM' 3 h2-3.8 FUEL LOADING AND REFUELING 3 kh 39 RADIOACTIVE MATERIALS.

'3-h6 L

3.10.

MISCELLANEOUS-RADIOACTIVE MATERIALS SOURCES 3-h6 3.11-HANDLING OF IRRADIATED FUEL'-

3-55 3.12

. REACTOR BUILDING POLAR CRANE 3-57 3.13 SECONDARY ~ SYSTEM ACTIVITY-3-58 3.14 FLOOD 3-59

' 3'.14.1 PERIODIC INSPECTION OF THE DIKES AROUND TMI 3 3.14 2

. FLOOD CONDITION FOR PLACING.THE UNIT IN HOT STANDBY 3-60

, 3.15 -

SHOCK SUPPRESSORS-(SNUBBERS)

.3-63 3.16 (RESERVED) 3.17 REACTOR BUILDING AIR TEMPERATURE' 80'

= 3.18'

-FIRE PROTECTION 3-86 2~

3.19 CONTAINMENT SYSTEMS 3-95 l

h" SURVEILLANCE STANDARDS k-1

'4.1 OPERATIONAL SAFETY REVIEW

.h-1 4.2 REACTOR COOLANT SYSTEM INSERVICE INSPECTION h-11 14'. 3 -

' TESTING FOLLOWING OPENING OF SYSTEM 1h-28 4.4 REACTOR BUILDING h-29

- k. b'. l' CONTAINMENT l LEAKAGE TESTS-

'h-29 4.h.2 STRUCTURAL INTEGRITY h-35 4 '.~ h'. 3L BYDRC ;EN PURGE SYSTEM h-37

,4.5-

/D4ERGENCY LOADING' SEQUENCE AND POWER TRANSFER, B4ERGENCY -

CORE COOLING SYSTE4-AND' REACTOR BUILDING COOLING FYSTD4 PERIODIC TESTING

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Jh.5.11 E4ERGEICY' LOADING SEQUENCE:

h-39 h.5 2[

EMERGENCY ' CORE. COOLING SYSTE4 h h1

-k.5.3:

REACTOR' BUILDING COOLING AND ' ISOLATION SYSTE4 h h3 4.5.h

. DECAY? HEAT REMOVAL SYSTE4 LEAKAGE'

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! LIST OF TABLES 1 r' -..

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iTablef

' Title Page 2.3-14

- Reactor Protection System Trip Setting Limits: 9

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'3.541:

Instruments operating Conditions 3-29' 3'.16-1.-

Safety Related Shock"Suppressors-(Snubbers)' 65' 3 18-1i

. Fire Detection Instruments-3-87_

1 h11-1 cInstrument Surveillance. Requirements-4-3 --

~h ~h.1-2 ~

Minimum' Equipment Test Frequency 4.1-3 ~

Minimum Sampling Frequency h-9 4.2 Instrument Surveillance Program h-lh' 4. 2-2 _-

Surveillance Capsule Insertion'& Withdrawal Schedule at.

4-27a TMI-2 4.h-1~

' Deleted i h. h -2

- Deleted 4.h-3 Deleted-h.15-1

_ Radioactive Liquid Waste' Sampling and Analysis h-59_ -

k.15-2~

Radioactive Gaseous Waste Sampling and Analysis h-63:

h.19-l' Minimum Number of Steam Generators to be Inspected

.. h-8h '

During Inservice Inspection

-h.19-2 Steam' Generator. Tube Inspection h-85 6.12 Deleted'-

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c FIGURESI-TITLE 13.5-2FJ Deleted.

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LLOCA Limited Maximum Allowable Linear Heat Rate.TMI-1

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f APSR Position Limits for. Operation' from 0 EFPD. to EOC

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'3 5-2I5 Deleted hs 3.5-2J1 Deleted.-

3. 5 -2K 1 Deleted i3 5-2L-Deleted 3 5-2M

-Deleted

. 3 5-2N

-Deleted.

3 5-3 Incore_ Instrumentation-Specification, TMI-1 h.2' -13 I

i Equipment and Piping Requiring Inservice Inspection in

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.Accordance with Section XI of the'ASME Code h. h Deleted h.h-2L Deleted-P

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- Deleted

= h. h--h '

-Deleted

- k. h-5.

Deleted-

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6-1 l

Met-Ed Corporate. Technical Support Staff'and j,

Station Organization Chart i

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. CONTAINMENT-SYSTEMS-4.'~

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CONTAINMENT STRUCTURAL INTEGRITY

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"M APPLICABILITY: ' Applies 'to the structural integrity of the reactor building.

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OBJECTIVE: ETo_ define the inservice tendon surveillance program for-the reactor building prestressing system.

Specification

~3.19 1.1

'With the structural integrity of the containment not conforming to' the; requirements of h.h.2.1.1.b, perform an-engineering evaluation of the structural integrity.of the containment to determine if COLD SHUTDOWN

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is' required. ;The margins'available in theLcontainment design may be considered during the investigation..If the acceptability of the: con-tainment tendons cannot be established within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, restore the structural integrity to within the 11mits within 2h hours or be in at-1

-least HOT STANDBY vithin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

3.19.1.2

_ With-the structural integrity of the containment otherwise not conforming to-the requirements of Specification 4.h.2.1,' submit a report _to the Commission pursuant to.NRC Regulatory Guide 1.16, Rev. 3.

This report shall include a description of the tendon condition,'the condition of the concrete (especially at tendon anchorages), the inspection procedure's, the tolerances on cracking, and the corrective actions taken.

3-95

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434.21 Structural'IntegrityL Specification' l h. k. 2.1:

Inservice Tendon Surveillance Requirements-n.c o _r-~

s TheLsurveillance' program for structural integrity and corrosion pro-tection conforms to the recommendations of the U.S. Atomic Energy -

Commission Regulatory Guide 1.35, proposed Revision 3. " Inservice -

Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures." The detailed surveillance program for the prestressing system tendons shall be based on periodic inspection and. mechanical tests. to fbe performed on selected tendons, as specified hereafter.

~h.h.2.1.1 -Containment-Tendons Tendon surveillance was completed for one year and three years following th'e initial structural' integrity using a Tech. Spec. based on Regulatory Guide 1.35 Rev. 1.

The containment tendons structural integrity shall be demonstrated at the end of five years following the initial-contain-ment structural integrity test and at five year intervals thereafter by:

Determining that for a representative sample" of at least 23 a.

tendons (6 dome, 7 vertical, and 10 hoop) each tendon has a lift off force equalling, or exceeding, its lower limit predicted for the time of the test as defined in NRC Regulatory Guide 1.35,

" Inservice Inspection for Ungrouted' Tendons in Prestressed Concrete Containments", Proposed Revision 3, April, 1979 If the lift off force of a selected tendon in a group lies between the prescribed lower limit and 90% of that limit, one tendon on each side of this tendon shall be checked for their lift off forces.

If the lift off forces of the adjacent tendons are equal to, or greater than,'their prescribed lower limits at the time of the test,-

the single deficiency shall be considered unique and acceptable.

If

'the lift off force of either of the adjacent tendons lies below the prescribed lover limit for that tendon, the condition is report-able per T.S. 6.9 2.A3 If the lift off force of any one tendon lies belov 90% of its pre-scribed lover limit, the tendon shall be considered a defective tendon.

It shall be completely detensioned and a determination made as to the cause of the occurrence. The condition is reportable per T.S. 6.9.2.A3 If the inspections performed at one, three, and five years indicate Eno. abnormal degradation of the post-tensioning system, the number of tendons checked for lift off force during subsequent tests may be 4

reduced to a representative sample of at least 11 tendons (3 dome,

-3 vertical, and 5 hoop).

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• For each inspection, the_ tendons shall be selected on a random but representative

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ibasis so that'the sample group vill change somewhat for each inspection; however, j

Lto develop a history of. tendon performance and to' correlate the observed data, one tendonL from each group.(dome, vertical, and hoop) may be kept unchanged after the initial selection.

h-35

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'YC 1b..Determiningthat[th'eaverageof.thenormalizedt.tendonlift.off.

forces for each tendon group (vertical, dome, and_ hoop) is equal to, or greater. than 1010. Kips for. vertical tendons,-1040 Kips for -dome "vvs~a

tendons,. and 1121 Kips for hoop tendons..If this requirement is

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'not met, the condition 'is reportable per 1T.S. 6.9 2. A3 and an =

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additional cample~ of h"., with 'a minimum of four and a maximum:of.

iten, of the; same group of. tendons shall be inspected.

If the total population 'of each group of ~ the sampled tendons meets the criteria

. above, the structural-integrity of the containment shall be considered acceptable.

Detensioning~one tendon in each group (dome, vert'ical and hoop) from c.

the representative sample.

One vire shall:be removed from each detensioned tendon and examined to' determine:

1. < That over the entire length of the. vire, the tendon wires ~ ) ave not undergone corrosion, cracks, or damage beyond that.which i

was originally recorded and the extent of corrosion is within specified acceptable limits.

Failure to satisfy these limits is a~ reportable ~ condition per T.S. 6.9 2. A3.

.2.

A minimum tensile strength value of 240,000 psi (guaranteed ultimate. strength of the tendon material) for at least three wire samples (one from each end and one at mid-length) cut from each removed wire.

Failure of any one of the wire samples to meet-the minimum tensile strength test is reportable per T.S. 6.9 2.A3 Upon retensioning, the elongation shall be within plus or minus 5%

of that. recorded at original stressing of the tendon.

If the 5%

limit is not met, an investigation shall be made to determine if wire failure is'the cause.

d.

Determining for each tendon in the above representative sample, that the sheathing filler grease is within acceptable limits as to:

1.

presence of voids.

2.

. presence of free water.

3 chemical and physical properties.

Failure of the grease to. meet acceptable limits is reportable per T.S.~ 6.9 2.A9 1

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• In ' order-for* the tendon lift ' off forces to be -indicative of. the average' level of l

?prestress, each lift off force is adjusted for differences which exist among the i

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tendons due to initial lock 'off-force and elastic shortening loss.

S h-35a c

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/ ih.k.2.1.2 f End Anchorages and Adjacent' Concrete Surfaces-

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The: structural 1ntegrityTof the-end anchorages of all tendons inspected 7

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, pursuant to-Specification 4.h.2.1.1 and-'the: adjacent concrete surfaces' shall be. determined through visualLinspection. The condition.of the end anchorage and adjacent. concrete shall? be recorded. The acceptance

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- criteria.shall be that all cra'ck vidths greater than 0.010 inch'shall

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,be recorded and' evaluated. Any crack vidth greater than 0.050 inch shall be cause.for investigation to determine the amount of structural impairment upon'the reactor building and!its continued structural

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integrity. Changes in the condition of the end anchorage or the concrete from.that previously recorded shall be noted on the record. Any condition'or change in condition which indicates abnormal material or structural behavior is reportable. -

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h.h.2J1.3 -Containment Surfaces

.The structural integrity of the exposed accessible interior and exterior surfaces of the containment, including the liner plate, shallfbe deter-mined during the shutdown for each Type A containment leakage rate test (specification h.h.1.1) by a visual inspection of these surfaces. This

' inspection shall-be performed prior to the Type A containment leakage rate' test. Any abnormal degradation must be documented and if it affects structural integrity it is reportable per T.S. 6.9 2.A9 i.

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4-36 (J

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[69[REPORTINGREQUIREMENTS;(cont'd);.

.x 6.9 3.~ Unique l Reporting' Requirements:

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L A. (Special: reports shall.be submitted to the Director of the o

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Office of-Inspection and= Enforcement Regional:-Office yithin l

lthe' time' period specified for each report. These reports shall be submitted covering the activities' identified below:

Tests Submittal Dates

(1) Containment Structural

' Integrity Test.

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(a).iTendon Surveillance Program' Within-3-months after i.

performance of surveillance

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. program.

(2). Steam Generator Tube' Inspection Within 3 months after L

' Program (See Section h.19 5) completion of inspection.

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(3) Containment Integrated Leak Within'6 months after

. Rate Test completion of test.

(4)

Inservice' Inspection Program Within 6 months.after U

five years of operation.

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FOOTNOTES 1.

A single.-submittal may be made for a multiple unit' station. The submittal.

should combine those sections that are common to all units at the station.

l 2.

This tabulation supplements the requirements 'of $20.h07 of 10 CFR Part 20.

i 3.

These reporting requirements apply only to Appendix A technical specifications.

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