ML20196J085

Proposed Tech Specs Implementing Option B of 10CFR50,App J & Allowing Frequency of Conducting ILRT & Local Leak Rate Testing to Be Based on Component Performance

ML20196J085
Person / Time
Site:	Fort Calhoun
Issue date:	07/25/1997
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20196J058	List: LIC-97-0124, Forwards Application for Amend to License DPR-40, Implementing Option B of 10CFR50 App J,To Allow Frequency of Conducting Integrated Leak Rate Testing & Local Rate Testing to Be Based on Component Performance ML20196J078 ML20196J085
References
NUDOCS 9708010291
Download: ML20196J085 (31)

Text

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TABLE OF CONTENTS (Continued) 1 EaEC 4.3 Nuclear Steam Supply System (NSSS).........

4-3 l

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 l

4.4 Fuel Storage.......................

. 4-4 4.4.1 New Fuel Storage

. 4-4 4.4.2 Spent Fuel Storage...

4-4 4.5 Seismic Design for Class l Systems...

. 4-5

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5.0 ADMINISTRATIVE CONTROLS

. 5-1 5.1 Responsibility....

. 5-1 5.2 Organi7.ation.........

. 5-1 l

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

...... 5-9 5.6 Reportable Event Action.

5.7 Safety Limit Violation

. 5-9 5.8 Procedures.....

. 5-9 1

5.9 Reporting 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 5.9.6 5-17a 5.10 Records Retention.

5-18 5.11 Radiation Protection Program..

5-19 5.12 DELETED 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...

5.17 Offsite Dose Calculation Manual (ODCM) 5-254 r

ss ntrol Program (PCP) 5-265 5;19 Containmem,.agitnestmg Programj ~, ^

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N5p26 6.0 JP

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. 6-1 4

6.1 Limits on Reactor Coolant Pump Operation 6-1 6.2 Use of a Spent Fuel Shipping Cask

. 6-1 i

l 6.3 Auxiliary Feedwater Automatic Initiation f etpoint 6-1 6.4 Operation With Less Than 75% of Incore Detector j

Strings Operable..

6-1 iin Amendment No. 32.34.43,54,55.57

• "s, OA 9

Of On ni AA 1Af ffi 1 f-T a

UV, UV. V /, / s, / /, a 'r &, a v &, A v i 9708010291 970725 PDR ADOCK 05000285 P

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l 3.0 SURVEILLANCE REQUIREMENTS l

l BASIS Specifications 3.0.1 through 3.0.4 establish the general requirements applicable to Surveillance Requirements. These requirements are based on the Surveillance l

Requirements stated in the Code of Federal Regulations,10 CFR 50.36(c)(3):

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" 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 l

Surveillance Requirements may be extended. It permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of l

plant operating conditions that may not be suitable for conducting the surveillance; l.

e.g., transient conditions or other ongoing surveillance or maintenance activities. It l

also provides flexibility to accommodate the length of a fuel cycle for surveillances that l

are performed at each refueling outage and are specified with an 18-month surveillance l

interval. It is not intended that this provision be used repeatedly as a convenience to l

extend surveillance intervals beyond that specified for surveillance that are not l

performed during refueling outages. The limitation of Specification 3.0.1 is based on i

engineermg judgement and the recognition that the most probable result of any paricular surveillance being performed is the verification of conformance with the Surveillance Requirements. 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 define the surveillance intervals for use in the Technical Specifications. This clarification is provided to ensure consistency in surveillance intervals throughout the Technical Specificaticns. 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 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. Unde te is

. 'on, t e

restrictive requirements of the ical S ifications ake preced ce over th s

sr ce er in. ;The requireme_nts;of regulations.take precedence over,the TSZTherefore test intervals governed by regulation cannot be extended _.by.the l

T 'Mn; example'of this~ exception is;the Containment leakage Rate Testing; Program l Specification 3.0.4 establishes a L vc in 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 l

requirements for the corresponding Limiting Condition for Operation. Under the provisions

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3-Ob Amendment No. 122,120,152 i

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3.0 SURVEILLANCE REOUIREMENTS 3.5 Containment Test Applicability l

Applies to containment leakage and structural integrity.

Ob_iective l

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 at:: 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.

A (2)

Containment SH&.g L;g ng; 7;;.; ro,,,,,,,,,,,

Perfoym;requiredjviisual: examinations and leakage rat (testinglin'faccordancepith;,the pontaimnent Ixakage Rate; Testing Programj q

(3)

Cantainment Penetratinnt I enk Data Tests (Tvoe B TestsYj Perform: required visualie.xaminations and leakage rate _ testing;in;accordance with the

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Containment Imakage, Rate; Testing Programjfor Ohe following penetrationsj t

(i)

Equipment Hatch (ii)

Personnel [AccegLock (iii)

M.echanical; Penetrations _M-1;through;M-99 (iv)

Fuel Transfer, Tube;(Mechanical Penetration M900)

(v)

Electrical Penetrationsj i.

j A41 D-9 D-6 F-2 E-HCV-383-3A l

A-2 B210 D-7 F-4 E-HCV-383-3B l

A-4

.B-) 1 D-8 F-5 E-HCV-383-4A l

A;5 Cs D-9 F-6 E-HCV-3834B I

l A'6 C-2 D 10 F,-7 Ai7 C4 D-11 F-8 A-8 C-5 E-1 F-9

&9 C-6 E-2 Ff10

3.0 SURVEILLANCE REOUIREMENTS 3.5 Containment Tests (Continued) 10 C-7 E-4 F-11 Akil 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 G4 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 Ib5 F,-1 H-4 i

(4)

Containment Isolation = Valves I ** Rate TestsVTvne C Tests)

Perform required' visual examinations'and leakage rate testing in accordancepith the q

Containment Leakage Rate Testing 1 Program for the following penetrations:

l M-2 M-31 M-52 IA'-3092 M-7 M-38 M-53 IA-3093 M-8 M-39 M-57 IA-3094 J

M-11 M-40 M-58 M-14 M M-69 M115 M43 M-73 M-18 M-44 M-74 M-19 M-45 M-79 M-20 M-46 M-80

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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 1D Tcs:s sha!! be condue:cd to assure ia: !cakagc of ec primary reac:or contammen; and associated sys:cas is maintaincd wiiin allowabic icskagc ra:c limi:3. Periodic suryci!!ance sha!! bc pcrformed to assurc proper main:cnance and Icak repair of ic containment structurc and pcac;ra:icn3 during ic plan:'s opcrating life.

Ocfim ions of term; used in the Icak raic testag specifica;;ons:

Lcakagc Rate - for :cs: purposes is ma: leakage of containmen; air which occurs in a uni: of timc. Sta:cd as a pcreentagc of wc;ght of ic original content of containment air l

a ic icakagc ra:c :cs: pressurc ia: cacapes to the outside atmosphcrc during a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> tes period.

M=!=m ^.!!c=S!: Ldage Rate 'L 1 - ic design bas icakagc rate of 0.1 % by weight ofic containmen; 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.

Ovcm!! In:cgra:cd Leakage Ratc - ma: Icakage ra:c which is obtaincd from a I

summation of Icakage ircugh all potential leakage pais including containment wclds, valves, fh:ings, and componcats which pcactraic containmen.

Acceptabic Critcria - :he standard against which tcs: results arc to bc compared for cs;ablishing ic functional acceptability of ic containment as a !cakagc limiting g

boundary.

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3-37 Amendment No. 95d5+

(Next Page is 3-45)

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3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment. Tests (Continued)

(3)-Integrated-LealtRate: Test:(TypeATest) a.

Introduction l

Type-A tests are intended-to measure the reactor containm t overall integrated leakage rate at periodic-intervals:

PretestRequirements A general-inspection of the-accessible-interioran exteriorsurfaces of the

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containment structures and components shall-b performed priorto any-Type-A j

st to uncoverany evidence of-structural de ioration-which may2ffect either th containment structural-integrity or-leak ightnessrIf there-is evidence of stru ural deteriorationrthe-Type-A-tests all not-be performed until corrective action ' taken in accordance with repa' proceduresrnon-destructive examina onsrand tests as specified-i the applicable code-specified in-10 CFR Part-50:5 at-the commencement repair workrSuch structural deterioration and-correcti actions taken-shal e reported as part-of the-Type-A--test seportr During the peri between th initiation of the-containment-inspection and performance of th Type-Aj estrno repairs or adjustments shall-be made so that the containment can e tested-in as close-to the'as-is" condition-as practicair During the period-bet n the completion of one-Type-A-test and the initiation of the containment-in e ion for the-subsequent-Type-A-testrrepairs or adjustments shall-b made-components-whose leakage exceeds-that-specified in the-Technical-ecificatio ras soon as practical after identificationrThis requirement-is' terpreted not preclude performance of-Type-B and Type C testing and te ired repairs prio o initiation of the containment-inspection and the perform ce of-the-Type-A-tes -

If durin a-Type-A-testrpotentially ex ssive leakage paths-are identified-which i

interf e-with satisfactory completion of e testror which result-in the-Type A test at meeting-the acceptance criteriarth Type-A-test shall be temporarily s pendedrThereafterrrepairs-and/or adjust ents to equipment-shall-be made d the-Type-A-test resumedrThe corrective-tion takenrthe change in leakage rate resulting-from the repairs and overa integrated-leakage determined j

from the Type-A and-local' leak rate tests-shall be i luded-in a report submitted to the Commissionc Closure of containment-isolation valves for the Type-A-tt t shall-be j

accomplished by normal operation and without any prelimi ry exercising or adjustments-(e.gtrno tightening of valve after closure by val -monitor)r Repairs of maloperating orleaking-valves shall be made necess y.-Information on any-valve closure malfunction orvalve leakage that requires-rrective action before the testrshall be included in the-Type-A Leak-Test-R ort submitted to the Commissionc 3-38 Amendment No. 95,151 J

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.0 SURVEILLANCEREQUIREMENTS 4,

3.

Containment _ Tests (Continued)

The contain nent-test conditions shall stabilize for a period of approxituate -4 hours priorto the start of the-leakagetate-testr i

Those portions of the-fluid systems-that-are part of the reactor cool t pressure boundary and-are open directly-to the containment-atmosphere-un r post accident conditions-and-become an extension of the bound y of the -

i containment shall be opened orvented to the containment-at sphere prior to l

and-during-the testrPortions of closedsystemsinside cont inment that penetrate l

ntainment-and rupture 2s a result of a-loss of coolant-a ident shall be-vented to e containment atmosphererAll-vented systems sh -be drained of water or othe uids-to the extent-necessary to-assure exposu of the system containment isolati valves to containment airtest pressure a to assure they will be subject o thepost-accident differential press rSystems that are required to maintain-th plant-in a safe condition during t -test shall-be operable in their normal mode, and need-not be-ventedrSys s-that are normally-filled-with waterand ope ing-underpost-accident-nditionsrsuch as-the containment heat -

. removal-system a :the component coo ng watersystemrneed not be-ventedr Howeverrthe conta ent-isolation ves-in--the systems-defined in-this section shall be-tested in acco ancevith-tion-3:5(5).-The measured-leakage rate from these-tests shall be eport o-the Commissionr c.

Test: Methods i

I All-Type-A-tests-shall-conduc. -in accordance-with-the provisions of 10 CFR Part-50rAppen

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The-accuracy of y-Test-A shall be v 'fied-by a supplemental-test:-The supplemental-t t method-selected-shall conducted-for sufficient-duration to establish acc tely-the change in-leakage r e between the-Type-A-test-and-the suppleme 1-Type-A-testrResults-from the plemental-test are acceptable provid he difference between the-supplemen I test data-and-the-Type-A-test data-i ithin 0:25-Lrlf-results are not-within 0. 5-L,rthe reason-shall be det inedrcorrective action-takenrand-a-successf supplemental test p formedr Test leakage rates-shall be calculated using absolute value corrected for instrument error:

d.

Acceptance Criteria The maximum allowable leakage rate shall-not exceed 0:1%

3-39 Amendment No.-95,1 i

3.

SURVEILLANCE REQUIREMENTS l

3.5 Containment _ Tests (Continued)

The total measured leakage rate at a pressure of 60 psig shall-be le. than 0.75-Lrlf local-leakage measurements are taken to effect repairs in-er to meet j

Oc75-Lcacceptcrce criteriarthese measurements shall-be taken-t a pressure of j

60 psig:

If two consecutive-Type-A-tests fail to meet the accepta e criteria; notwithstanding the requirements of-the testing-freque cyra Type-A-test-shall be rformed on a-refueling-frequency-until two conse tive-Type-A-tests meet the 1

ac tance-criteriarafter which time the normal-t ting frequency schedule may be r umed e.

Testing. equency A-set of-thre Type-A-tests shall be pe ormedrat approximately equal-intervals during each-10 ' ear servicercriodr he-third test of each set-shall be conducted when the plant-is hutdown-for th 0-year-in-service inspections:-

The performance of-pe-A-t ts-shall be-limited to periods when the plant facility-is non-operatio l-d secured-in-the shutdown condition under administrative control an in accordance-with the safety procedures-defined in the licenser (4)

Containment: Penetrations eak-Rate:

ts.(Type:B. Tests) a.

Introduction Type-B-te s are-intended-to detect-loca eaks and-to measure leakage across each pr sure-containing orleakage limiti -boundary-for the containment pene tions b.

st: Methods Type-B-tests-shall-be performed by local pneumatic essurization of the l

containment penetrationsr either individually or-in gro srat a pressure of 60 psig -

Examination shall-be performed by halide-leak-detection me od or by-other equivalent test methods such as measurement-of-the rate of ma up required to maintain the test volume at 60 psig f

3-40 Amendment No. 95;1 1;157 l

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<3 SURVEILLANCE REQU;2EMENTS l

3.

ContainmenLTests (Continued) l l

c.

Acceptance Criteria l

l The combined-leakage rate of-all penetrations-and valves subject t ype-B and Type Gtests shall be less-than-or equal to 0.6-L,:

l If at-any time it-is-determined that-a-leakage rate is-greater-an 0.6 L,rrepairs i

shall be initiated-immediately-If repairs 1tre not complet and-conformance to, the-acceptance-criteria-is not demonstratedwithin -48 h rsrthe reactor shall be ut-down and-depressurized until repairs are compi d and-the-local-leakage m ts-this1icceptance criteria:

i-l The-r its ofpersonnel access-locle door seal sts at-5 psig shall-not exceed,01 L,:

d.

Testingfre ency Type B tests sha beperformed du g each refueling outageror other convenient-interva rbut-in no c at intervals greater than-2 yearsrexcept the personnel access-loc -(PAL) w ch will be-tested as-follows:

l (i)'

Every six1nonth t entire PAL assembly shall-be-leak tested at-60 psig:

(ii)

If-the PAL-i pened-ing periods-when containment-integrity-is-not requiredr

-PAb door-alsshall-be lealetested at-5 psig at the end-of such per' s1md-the entire AL-assembly-shall then be leak-tested at 60 psig-w' in-two-weeks of ac ving-the required condition-for con inment-integrityr (iii)- -the-PAbis opened during thein val between the six-month tests when containment-integrity 1s requir rthe-PAbdoor seals shall be-leak tested at a pressure not-less than-5 psig ithin-72-hours -If the PAL-is l

opened more frequently-than once per-7 oursrthe door seals shall be leak tested at-a pressure of-5 psig at-least o ce every 72-hours during the period of-frequent openings: '

Penetrations.to.be-Tested.m (i)

Equipment Hatch l

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

(v)

Electrical-Penetrations 3-41 Amendment No.-95

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.0 SURVEILLANCE. REQUIREMENTS 3.

Containmentlests (Continued) i A 1 --

B-9 D-6

-F2 E HCV 383 3A A2 B 10 D-7 F-4 E HCV 383 3B A4 B ll D8 F5 E-HCV 383 4A

-A 5 C1 D9 F6 E HCV 383 4B l

-A6 C2 D 10 F7

-A 7 C4 D Il F-8 8

C5 E1 F9 A

C6 E2 F 10 A1 C-7 E4 F 11 A ll-C8 E5 G1 l

B1-C9 E-6 G2 i

l B2 C-10.

E7 G-3 B-4

-G ll E-8 G4 B5-

- D-1 E-9 H+1 B-6

-D 2 E 10 H2 i

B7 4

E 11 H3 B8 F1

-H 4 (5)

Containment-Isolati Nalvesleak-Ratelests.(T C-Tests) l l

a.

Introduction l

Type C-tests are-int ded to me re containment isolation-valve-leakage ratesr b.

Test-Methods l-Type Otests shall-be-for ed by-local pressurization with air or nitrogen at a pressure of 60psign hepres re-shall be applied-in-the same direction-as that L

when thevalve wo d berequir -to perform its safety-functionrunless-it can be determined-that-eresults-from-tests-for apressure applied-in a-different direction will rovide equivalent or-ore conservative resultsrEach valve-to be tested shall-

' closed by normal opera 'on and without any preliminary exercisin or1idjustments-(e.grrno tight ing of valve aftertlosure by-valve motor) l c.

A eptance_ Criteria The combined-leakage rate of all penetrations and 'alves subject-to-Type-B and Type Gtests-shall-be less than or-equal-to 0.6-1 r r-the purge isolation-valve 3

i.

testsrthe measured purge valve-leakage rate shall be s bstituted for the purge valve leakage rate from the last-complete-Type-B and C st-and the total leak rate recomputedr Leakage of the containment air purge isolation valves shall no exceed-18,000 standard cubic centimeters per minute (SCCM)rIf the leakage te is determined to be greater than 18,000 SCCMrrepairs shall be-init ted 3

immediately-in order to meet-this acceptance criterion.-

t 3-42 Amendment No. 68,95;151

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l 3.0 SURVEILLANCEREQUIREMENTS

.5 Containment. Tests (Continued)

.1 l

If at any time it-is determined that a-leakage rate-is greater than-0.6 Lc epairs

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shall be-initiated-immediatelyrlf repairs are not completed 2nd-conf ance to

, the acceptance criteria-is not-demonstrated within-48 hoursrthe re torshall-be shut down and depressurized until repairs are completed and the ocal-leakage meets-this acceptance criteriar

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--Testingfrequency j'

Type C-tests shall be performed-during each refuelin outageror other nvenient-intervalsrbut-in-no case at-intervals gre er than-2 yearsrThe e tainment-purge-isolationvalves shall21so be-akage testedpriorto bringing the-actorout of each cold-orrefueling shutd but-in no casemt intervals

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greate than nine monthsrlf the purge valv are-opened during cold or l

refuelin hutdownrthe lealetest2 hall be-rformed afterthe purge valves are closed-for e-last-timer e.

Penetrations:t

-Testedm l

M2 M-52 IA-3092 M-7 M--3 8-

-53 IA-3093 M8' M 39- - -M 57 IA-3094 M-1-1 M40

- M 58 M 14 M -M-69 l

M 15 M4

-73 l

M 18 M4 4

M 19 45 M+

M 20 M46 M 80 M-22

-M 47 M-87 7

- M 24 !

M-48 M-88 M-25 M-50 M HGV 38 M3 M-51 M HCV-383-(6)-Specia estingRequirements A

majormodification or replacement of a-component-wh

-is part of the -

ntainment boundary shall be-followed by-either-Type-art e-Bror-Type G tests 2s applicable-for the area 2ffected by-the modification-and shall m t-the applicable acceptance criteria:-Minormodificationsror replacementsrperfo ed directly prior to the conduct of a scheduled-Type-A-test-do not-require 2-separate-tes l

(7)-Report:on:TestResults Lealerate-tests-shall be the subject of2 summary technical report submitted the l

Commission approximately-3-months ~after the conduct of each test.-The repo shall be L

titled' Reactor Containment-Building Integratedteak-Rate-Test;'

3-43 Amendment Noc24,95;151 i

3.0 SURVEILLANCE.. REQUIREMENTS 3.5 Containment _ Tests (Continued)

The report shall-contain an analysismnd-interpretation of the-Type--A-test resul and a summary analysis of periodic-Type-B and-Type G tests that were performed mee the last-Type-A-test:

Leakage test results-from-Type-ArBrand Otests that-failed-to meet-t applicable acceptance criteria shall-be reported-in-a separate summary report-proximately-3 months after the conduct of-these-tests.-The-Type-A-test report s all-include an alysis-and interpretation of the-test data,-the least squares-fit nalysis of the-test data (T

-A-tests only),-the-instrumentation error analysis-(Typ A-tests-only) rand-the i

struc ral conditions of thecontainment or componentsrif nyrwhich-contributed to-the failure ' meeting-the acceptance criteria.-Results and a lyses of the-supplemental verificati -test employed to demonstrate the validity-the-leakage rate test measurem ts shallalso be-included:

4 3-44 Amendrnent No.-95,97

3.0

' SURVFII I.ANCE REOUIREMENTS 3.5 Containment Tests (Continued).

(8)(5) rveillance for Prestressing System J

Samnle Selection a

The 210 dome tendons and 616 helical wall tendons shall be periodically inspected for r

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.

i b.

Visual Insnection The following visual inspections shall be performed:

(i)

The exterior surface of the containment sha!! be visually examined to detect areas of large spall, severe scaling, D< racking 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 w' 3.5(8)(5)a. the I

tendon anchorage assembly hardware shall be visually in,pd _

igns of abnormal material behavior or wear.

1 (iii)

The cencrete surrounding the visually inspected tendon anchorages shall be i

visually inspected for signs of sig.cficant 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 of the grease caps is not necessary for this inspection.

c.

Prestress Monitoring Tests j

Liftoff tests shall be performed on each tendon seleti 'i in accordance wi 3.5(8)(5)a.

l to monitor prestress. Additionally, the tests shall include the followmg:

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3-45 Amendment No. 95,97,139,151 J

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I 3.0 SURVEII.I.ANCE REOUIREMENTS 3.5 containment Tests (Continued)

The tendons detensioned in accordance wi 3.5(8)(5)c.

may be the tendons from which the sample wires are removed The co tr tendons shall NOT be included as tendons to be detensioned or have s removed. In addition, all wires found to be broken shall be removed for tensile testing and visual examination.

e.

Inspection 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 9,

" 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 che containment structure, f.

Acceptance Criteria (i)

No evidence of nifican etur eteriora 'on of the concrete inspected in accordance w 3.5(8)(5)b.(

a 3.5(8)(5)b. ii) which may affect the structural inte ty o the itai nt st c re can be detected.

3-47 A.m.:ndment No. 95,97,130,151

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 significtmt, 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 1

potential reduction of Containment structural integrity. An investigation shall

}

also be made to determine which tendons could have lost the greese and whether the gra se loss has adversely affected their corrosion protection.

st eh force measured for each tendon liftoff' tested in accordance (ii)

The wi i 3.5(8)(5)c.Shall be compared with the limits predicted by USAR Fig 5.10-

[

3.

f the m ed prestressing force of a selected tendon is greater than the prescri ower 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.

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

3.0 SURVEIIIANCE 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 sr

'the first tendon, shall be liftoff tested. If the prestressing forces of c

. the second and third tendons are greater than 95% of the prescribed lowu amit, 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, lift '

sss be performed on additional tendons to determine the cause d extent of su h occurrence. This occurrence shall be considered reportable r 3.5@(5)g. tf the measured prestressing force of a selected tendon is I ss than 90%

the prescribed lower efective tendon shall be fully ins etermine the cause and extent of su occurrence. This occurrence shall be considered reportable per 3.5@(5)g.

average prestressing force of all measured tendons of a group (corrected for average condition) is found to be less than the prescribed lower limipn investigation shall be per'ormed to determine the cause and exten of such an occurrence. Such an occurrence shall be considered reportable r 3.5@(5)g.

If from consecutive surveillances the average measured prestressing eo 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 ad ionalliftoff tests shall be performed to determine the cause and exten f s ch urrence. Such an occurrence shall be considered reportable per.5@(5)g.

r (iii) f during th etensia.ing and retensioning of.

m accordance with 3.5(e)(5)c., t e elongation corresponding to a specific load differs by more than 10 % o at rem ded during installation of the tendons, an investigation shall to e difference is not related to wire failures or slippage of wires in chorages.

'fference of more than 10% shall be considered reportabl er 3.5@(5)g.

1 3-49 Amendment No. 95,97,13?,l '+

l

~.

3.0 SURVEIIIANCE REOUIREMENTS 3.5 Containment Tesis (Continued)

(iv)

The minimum acceptable ultimate tensile strength of the wire samples to be tensile tested shall be 240,000 si with a minimum elongation of 4%

in accordance with ASTM A

-5 pe BA wire. Failure in the tensile test at strength or el ngation valu less than those specified shall be considered ble r 3.5(8X5)g.

ther conditions which indicate corrosion f by vis inati

. the wire shall be considered reportable r 3.5(8)(5)g.

(v)

Results of the 1 ratory tests and examinations of the filler grease will be considered acceptable if the following conditions are raet:

(a)

Water content i10% by weight

]

(b)

Chlorides A 10 ppm (c)

Nitrates A 10 ppm (d)

Sulfides A 10 ppm (e)

Reserve alkalinity

>0 (Base numbers)

(f)

The difference between the amount of grease injected into a j

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 e abov itions for the filler grease shall be considered reportar per 3.5(6)(5)g.

I g.

Corrective Action and Reoorti 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.

i 1

i

[

3-50 Amendment No. 24,5S,05,139,151 l

l 3.0 SURVEH LANCE REOUIREMENTS l

3.5 Containment Tests (Continued)

L h.

Test Freauency l

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

Basis l

l The containment is des 2gned 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 l

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 l

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 l

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 i per (o

cool'

't) ublic osure i

d w

belo CF rt 100 values in the event of t maximum l

hypothetical accident.m The performance of a pcrisdic an integrated leakage rate test and_ performance oflocal leak rat _e testing ofindividual penetrations,at periodic intervals during plant life provides a current assessment of potential lea age from the containment.

l 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 tes ssure tends to unse he ilie

%C 1 c.

c en pressure t s

eat the resilient seals. The six month A periodic test ensures the overall PAL integrity at 60 psig.

The ficqucacy of thc pcriodic integrated leakage rate test (Type A test) is kcycd to the refac ing schcdulc for thc scactor, because this :cs can only be performed during refueling shutdowns.

i 3-51 Amendment No. 53,97,13,151

3.0 SURVEILLANCE REOUIREMENTS 3.5 Containment Tests (continued)

[The specified equency of perio Ped 1=1= r ts is based on three several r jor considerati su First is ih[(1,) There isja w probability of leaks i t

.er because of the the

-TIgumess%r me welds during i

(

n and conformance of the comp e containment to a low leak rate at 60 psig during pre-operational testing, which is consistent with 0.1 % leakage at design basis accident conditiopigph%g$

- - ng w operatior. Second is the morgfrgquenti(2)iPeriodic testing; l

1s;; conducted at the ull acciden

~

s on ose po o j

l MAMac ope that are most likely t ks during reactor ation (penetr '

and isolation valvq j and t..c A w value (0.60LJ of l

th ta k

mt 's cified as accepthJC'/.m.trations mi isolation ves. Thirpjsjhc;(g endon stress surveillance prograg,j,hicQovides assuranceIngtan2nmuai art t

structur int rit of t ' o alnment is l

'n

(41Alevi

,0 eakage r _ s o tai.

'd, ring ast con Tent niegrated leakage rate testing is;copducted to' set appropriate; frequency of l

performance not to; exceed lonce every;10. years;l(5)]isuallinspecdon of,the j

containment structure is conducted ;every;otherfrefueling.andlprioritoleach Integrated. Leakage. Rate. Test; i

l As left leakage prior to theifirst startup after, performing a; required l_eakage test l

is; required _to; beg 0j6 Lifor combinediTypelBiand C; leakage,(and 150.75 Ljfor'overall Type A leakageE'Atlall other times between' required l

k leakage rate tests,ithe acceptance criterialis based'on an~overall Type Alleakage 1

l l

limit,of;s:ll.0 OJCAtfl.0 Eithejoffsite dose l consequences'are bounded by the l

assumptions of the' safety analysiss l

l Integrity tests of the purge isolation valves are established to identify excessive degradation of the resilient seats of these valves. Simultaneous testing of l

redundant purge valves from a leak test connection accessible from outside l

containment provides adequate testing. The testing method is identical to the l

Type C purge isolation valve test performed in accordance with 10 CFR Part i

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 prestressing system. Allowances have been made in the reactor building design for the reduction and changes. Through 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.

I i

3-52 Amendment No. 97,B9

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e 3-53a Amendment No. %

5.0 ADMINISTRATIVE CONTROLS l

5.9.3 Snecial Reports Special reports shall be submitted to the Regional Administrator of the appropriate NRC 1

Regional Office within the time period specified for each report. These reports shall be i

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.

l c.

Cont '

en ructu al tes r

enc-pDELETEDContainment cak ra:c cs:s, refcccace 2.5.

cial m mtena ce re rts.

d.

l l

e.

l f.

DELY ED g.

eria adia non su c.

anc im e

s, re erence 3.3.

l h.

DELETED l

i.

Post-accident monitoring instrumentation, reference 2.21 j.

Electrical systems, reference 2.7(2).

l 5.9.4 Unique Reporting Requirements a.

Annual Radioactive Effluent Release Report l

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.

l b.

Annual Radiological Environmental Operating Report i

i 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 l

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 j

(2)Section IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR 50.

l l

c.

Fire Protection Program Deficiency Report 1

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. 9,24,38,46,85,110, (Next page is 5-17a) 113,123,147,152,160,164 4

+=m W

DMINISTRATIVR CONTROLS 5;19 CaatainmenTil amiraeciate Testino lhogram I

Kprogramf.shal1 be established t.o; implement the leakaggrateitesting of,the_ containment as i

required by110_gFR:50.54(o)land 10 CFR 50,7 Appendix Jfdption;Bf as modified.by approved 3l 3

exceptionsGThis program shall be;in accordance with thel guidelines lpontainedLin Regulatory l

Guide)1l;163k"_ Performance-Based ContainmentLeak-Test. Program [ dated September ljp5;;

s l

as modified by,the;following exceptions; (1)

If the PersonneljAirl Lock-(PAL)lis opened _during periods;when containment integrity is notjrequired;the PAKdoor seals shall. be; te.sted;at the endfof suchliEriQQd the j

entire PAL shall be tested within;14; days efter RCS.ltemperatureLQ?210*Fj i

(2)

For;eachlPAIJdoor,; seal leakage rate isiO;01 Liwhen pressurized tof 5.0 psig(

(3)

For eagh;Coctainmentl Purge Valve; leakage ratejsM18,000.secmphen tested:at;2]

J Pd (4)

Iffat'any time;when containment integrity is required and.the total Type,B and Q measured leakage rate exceeds,0.60 LiMinimum. Pathway l Leakage ~ Rate,(MNPLR)]

repairs ~shall-be ini,tiated;immediately :If. repairs and retesting fail.to demonstr_ ate conformance to this; acceptance criteria.withinf8.hoursi then containment shall be declared inoperable:

The containment; design accident; pressure-(P,),.is 60 psigi l

l The' maximum allowable primary; containment leakage rate;;Ll,;at.P.Eshall.beLO.1% ofj containment air.we_ight per' day!

i Leakage' Rate acceptance; criteria ~are:

I ai

' Containment l leakage rate acceptance criterion isfs 1.0 Li.1During unit startup following testing in accordance.w.ith this. programf the leakage rate acceptance _ criteria i

are;s 0.60 L/ Maximum Pathway Leakage Rate,(MXPLR) for Type,B;and C tests.and

's 0,75 L ifor Type A tests.'

'l b;

Personnel / Air < Lock testing acceptance; criteria ~are; 1

i l

(1)

Overall-Pergonnel Air Lock _ leakage is s 0.1TLiwhen; tested at EPJ (2)

Fo_r each PAL; door, sealleakage rate is s 0.01 Liwhen pressurized to;2 5.0 psig; l

I c;

Containment Purge Valve (PCV-7.42A/B/C/D) testing acceptance criterion is:

For each Containment Purge Valve, leakage rate isl < 18,000 secm when~ tested at 23d TheMrovisions of_ Specification;3.0.1;do;not applf,..to the test frequenciesispecilled in the i

l Containment Leakage Rate Testing Program)

The p.rovisions of Specification 3.0.4 are applicable to the Containment Leakage Ratd. Testing j

Program.?

"^

~

n; Amendment 5-261 _

l U.S. Nuclear Regulatory Commission LIC-97-0124 A

AC V E \\~~

3 l

l t

I i

f

l l

l DISCUSSION, JUSTIFICATION AND N0 SIGNIFICANT HAZARDS CONSIDERATION DISCUSSION AND JUSTIFICATION:

The Omaha Public Power District (OPPD) proposes to revise the Fort Calhoun Station (FCS) Unit No.1 Technical Specifications (TS) to implement Option B of the recently revised 10 CFR Part 50 Appendix J.

The proposed implementation is consistent with Regulatory Guide 1.163, " Performance-Based Containment Leak Test Program," September 1995, industra guidance contained in NEI 94-01, Revision 0, " Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J." and NEI 94-01 Errata Sheet, and ANSI /ANS 56.8-1994, with the following clarifications and exemptions:

(1)

Section 9.2.4 of NEI-94-01 on Containment Repairs and Modifications specifies that Type A testing may be deferred for welds attaching to the steel pressure retaining boundary whose nominal diameter does not exceed one inch.

It should also allow Type A testing to be deferred for penetrations of the steel retaining boundary where the diameter does not exceed one inch.

(2)

Section 9.2.3 specifies at least a 24 month interval between Type A tests is required to establish a performance baseline.

It should state 24 months or refueling interval between Type A tests, since FCS is on an 18 month refueling interval.

In addition, OPPD requests that the following previously approved exemptions remain in effect:

(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 Tm > 210 F.

(2)

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

(3)

For each Containment Purge Valve, leakage rate is < 18,000 sccm when tested at 2 P,.

(4)

If at any time when containment integrity is required, the total Type B and C measured leakage rate exceeds 0.60 L., repairs shall be initiated immediately.

If repairs and retesting fail to demonstrate conformance to the 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.

1 a

i j

l DISCUSSION AND JUSTIFICATION (Continued):

The proposed change will revise TS 3.5(2) to reference the Centaltniient Leakage Rate Testing Program, add a new TS 5.19. delete TS 3.5(3)'through 3.5(7).

i revise the Basis.of TS 3.0 to state that regulations take precedence over TS.

and revise the Table of Contents. The requirements of 10 CFR Part 50 Appendix J (Type A. B and C-testing, and reporting requirements) that are currently addressed by TS 3.5(3) through 3.5(7) will be addressed by the Leak Rate

. Program incorporated into TS 5.19.

The proposed change potentially affects the leak-tight integrity of the containment structure designed to mitigate the consequences of a Loss-of-Coolant Accident (LOCA). The function of the containment is to maintain functional integrity during and following the peak transient pressures and temperatures and limit fission product leakage following the design basis LOCA.

Because the proposed change does not alter the plant design. only the frequency of measuring Types A. B and C leakage, the proposed change does not directly result in an increase in containment leakage.

Test intervals will be established based on the performance history of i

components being tested.

OPPD expects that the frequency of testing most containment isolation valves and penetrations will decrease based on a history of good performance.

However, the frequency of monitoring the relatively few j

containment isolation valves and/or containment penetrations subject to above normal leakage will not decrease. Thus, the probability that an increase in

_1 l

containment leakage could go undetected for an extended period of time is not l-increased since a performance based program will identify those valves and penetrations which must continue to be tested each refueling outage.

The risk resulting from the proposed changes is characterized as follows.

based primarily on the results contained in NUREG-1493 " Performance-Based Containment Leakage Test Program." the principal Technical Support Document used by the NRC as the basis for the Appendix J Final Rule:

Type A Testing NUREG-1493 found that the effect of containment leakage on overall accident risk is minimal since risk is dominated by accident sequences that result in failure or bypass of the containment.

Industry wide. Integrated Leak Rate Tests (ILRTs) have only found a small fraction of the leaks that exceed 4

current acceptance criteria. Only three percent of all leaks are detectable only by ILRTs. and therefore, by extending the Type A testing intervals, only l

.three percent of all leaks have a potential for remaining undetected for j

longer periods of time.

2 A

I l

l l

i DISCUSSION AND JUSTIFICATION (Continued):

In addition, when leakage has been detected by ILRTs. the leakage rate has been only marginally above existing requirementsr The Fort Calhoun Station Unit No.1 Type A testing confirms the industry-wide experience that a majority of the leakage experienced during Type A testing is through components tested by Type B and C tests.

These observations, together with the insensitivity of reactor accident risk to the containment leakage rate, show that increasing the Type A leakage test intarvals would have a minimal impact on public risk.

Type B and C Testing NUREG-1493 Dund that while Type B and C tests can identify the vast majority (greater thEn 95 percent) of all potential leakage paths, performance-based alternatives to current local leakage-testing requirements are feasible without significant risk impacts.

The risk model used in NUREG-1493 suggests that the number of components tested would be reduced by about 60 percent with less than a three-fold increase in the incremental risk due to containment leakage.

Since, under existing requirements, leakage contributes less than 0.1 percent of overall accident risk the overall impact is very small.

In addition, the NRC's Final Regulatory Impact Analysis concluded that while the extended testing intervals for Type R and C tests led to minor increases in l

potential off site dose consequences, the beneficial expected decrease in onsite worker dose received during ILRT and local leak rate testing exceeds (by at least an order of magnitude) the potential off-site dose consequences.

The containment isolation system is designed to limit leakage to L., which is stated in the Fort Calhoun Station Unit No.1 Technical Specifications to be 0.1 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 60 psig.

The limitation on containment leakage rate is designed to ensure that total leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure.

The L, value is not being modified by this proposed change.

l l

i 3

BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION:

l l

The proposed changes do not involve significant hazards consideration because operation of Fort Calhoun Station Unit No. 1 in accordance with these changes would not:

(1)

Involve a significant increase in the probability or consequences of an j

l accident previously evaluated.

j The proposed change implements Option B of 10 CFR Part 50 Appendix J on performance-based containment leakage testing. The proposed change does not involve a change to the plant design or operation.

As a result, the proposed change does not affect any parameters or conditions that contribute to the initiation of any accidents previously evaluated.

The proposed change potentially affects the leak-tight integrity of the containment structure designed to mitigate the consequences of a Loss-of-Coolant Accident (LOCA). The function of the containment is to maintain functional integrity during and following the peak transient pressures and temperatures and limit fission product leakage following the design basis LOCA.

Because the proposed change does not alter the plant design, only the frequency of measuring Type A. B. and C leakage.

the proposed change does not directly result in an increase in containment leakage.

Test intervals will be established based on the performance history of compone:1ts being tested. The frequency of monitoring the relatively few t

l containment isolation valves and/or containment penetrations subject to above normal leakage will not decrease by implementing Option B of Appendix J.

A performance based program will identify those valves and penetrations which must continue to be tested each refueling outage.

The risk. resulting from the proposed changes is characterized as follows, based primarily on the results contained in NUREG-1493

" Performance-Based Containment Leakage Test Program." the principal Technical Support Document used by the NRC as the basis for the Appendix J Final Rule:

1 4

a

5

' BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION (Continued):

Type A Testing NUREG-1493 found that the effect of containment leakage on.overall accident risk is minimal since risk is dominated by accident sequences that result in failure or bypass of the containment.

Industry wide, Integrated Leak Rate Tests (ILRTs) have only found a ~ small fraction of-the leaks that exceed current acceptance criteria.

Only three percent t

of all leaks are detectable only by ILRTs, and therefore, by extending the Type A testing intervals, only three percent of all leaksLhave a potential for remaining undetected for longer periods of time.

In addition, when leakage has been detected by ILRTs, the leakage rate has been only marginally above existing requirements.

The Fort Calhoun Station Unit No.1 Type A testing confirms the industry-wide experience i

that a majority r, the leakage experienced during Type A testing is through components tested by Type B and C tests.

NUREG-1493 found that these observations, together with the insensitivity of reactor accident risk to the containment leakage rate, show that increasing the Type A leakage test intervals would have a i

minimal impact on'public risk.

Type B and C Testing NUREG-1493 found that while Type B and C tests can identify the vast l

majority (greater than 95 percent) of all potential leakage paths, performance-based alternatives to current local leakage-testing requirements are feasible without significant risk impacts. The risk-

~

l~

model used in NUREG-1493 suggests that the number of components tested i

would be reduced by about 60 percent with less thar,e three-fold increase in the incremental risk due to containment leakage.

Since.

under existing requirements, leakage contributes less than 0.1 percent of overall. accident risk, the overall impact is very small.

In addition, the NRC's Final Regulatory Impact Analysis concluded that while the extended testing intervals for Type B and C tests led to minor increases in potential offsite dose consequences, the beneficial

-expected decrease in onsite worker dose received during ILRT and local l

leak rate' testing exceeds (by at least an order of magnitude) the potential off-site dose consequences.

Therefore, the proposed change will not result in a significant increase in the probability or consequences of any accident previously evaluated.

I 5

a

._.m

..m..

BASIS FOR N0 SIGNIFICANT HAZARDS CONSIDERATION (Continued):

(2)

Create the possibility of a new or different kind of accident from any accident previously evaluated.

There will be no physical alterations to the plant configuration, changes to setpoint values, or changes to the implementation of setpoints or limits as a result of this proposed change.

As a result, the proposed change does not affect any of the parameters or conditions that could contribute to initiation of any accidents.

This change involves the reduction of Type A. B, and C test frequency.

l Except for the method of defining the test frequency, the methods for l

performing the actual tests are not changed.

No new accident modes are created by extending the testing intervals.

No safety-related equipment or safety functions are altered as a result of this change.

Extending the test frequency has no influence on, nor does it contribute to, the possibility of a new or different kind of accident or malfunction from those previously analyzed. Therefore, the proposed change does not create the possibility of a new or different kind of xcident from any previously evaluated.

(3)

Involve a significant reduction in a margin of safety.

The proposed change only affects the frequency of Type A, B, and C testing.

Except for the method of defining the test frequency, the methods for performing the actual tests are not changed.

The frequency of monitoring the relatively few containment isolation valves and/or containment penetrations subject to above normal leakage will not decrease by implementing Option B of Appendix J.

A performance based program will identify those valves and penetrations which must continue to be tested each refueling outage.

NUREG-1493 has determined that, under several different accident scenarios, the increased risk of radioactivity release from containment is negligible with the implementation of these proposed changes.

The margin of safety that has the potential of being impacted by the proposed change involves the offsite dose consequences of postulated accidents which are directly related to containment leakage rate.

The containment isolation system is designed to limit leakage to L, which is stated in the Fort Calhoun Station Unit No. 1 Technical Specifications to be 0.1 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 60 psig.

6 1

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BASIS FOR N0 SIGNIFICANT HAZARDS CONSIDERATION (Continued):

The limitation on containment leakage rate is designed to ensure that total leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure.

The margin to safety for the offsite dose consequences of postulated accidents directly related to the containment leakage rate is maintained by meeting the 1.0 L.

acceptance criteria.

The L, value is not being modified by this proposed change.

Except for the method of defining the test frequency, no change in the method of testing is being proposed. The Type B and C tests will continue to be done at 60 psig or greater. Other programs are in place to ensure that proper maintenance and repairs are performed during the service life of the primary containment and systems and components penetrating the primary containment.

Therefore, the proposed change will not result in a significant reduction in a margin of safety.

Based on the above considerations, it is OPPD's position that this proposed amendment does not involve significant hazards considerations as defined by 10 CFR 50.92 and the proposed changes will not result in a condition which significantly 9ters the impact of the Station on the environment.

Thus, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and pursuant to 10 CFR 51.22(b) no environmental assessment need be prepared.

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U.S. Nuclear Regulatory Commission LIC-97-0124 A

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l MILESTONE SCHEDULE FOR IMPLEMENTATION O_E OPTION B TO APPENDIX J AT FORT CALHOUN i

STATION SCHEDULE ACTUAL EVENT 1-17-97 1-31-97 Review industry information, other plant's implementation, NEl Guidelines, ANSI /ANS 56.8, and Regulatory Guide 1.163 4-7-97 3-14-97 Develop guidelines to assess performance data.

9-26-97 Complete evaluation of Integrated Leak Rate Test (ILRT) performance data.

10-31-97 Complete evaluation of Local Leak Rate Test (LLRT) performance data.

11-28-97 Complete common mode failure analysis.

11-28-97 Complete review of Probabilistic Risk Analysis (PRA) data for all tested penetrations.

12-19-97 Complete Inservice Testing (IST) interface effect analysis.

1-16-98 Develop revision to Program Basis Document (PBD).

1-30-98 Develop required proc = dure changes.

3-2-98 Receive Amendment.

3-3-98 Submit PBD revision to management for approval and I

issue.

3-3-98 Submit required procedure changes for review and signature.

l 4-1-98 Implement Option B (Type B and C tests) for 1998 l

outage.

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