ML20137H609

Proposed Tech Specs Re Primary Coolant Specific Activity

ML20137H609
Person / Time
Site:	Farley
Issue date:	03/25/1997
From:	SOUTHERN NUCLEAR OPERATING CO.
To:
Shared Package
ML20137H605	List: ML20137H601 ML20137H609
References
GL-95-05, GL-95-5, NUDOCS 9704020296
Download: ML20137H609 (16)

Text

__.. _ _ -. _ _ _.,

3/4.4.9 SPECIFIC ACTIVITY a

l LIMITING CONDITION FOR CPERATION i

f 3.4.9 The specific acttvity of the prim'y coolant shall be limited to:

Lees than er equal to derecurie per gram Dost a.

EQUIVALast? 1-131:

Less than er equal to 100/k microcurie per gram.

i h.

l APPLICABILITY:

M005.8 1,

• 2, 3, 4, and 5 ACT!cel:

}

0. 3 M00ts 1, 2, and 3*:

a.

With t cific eetivity of the primary coolant greater than +r481dicrocurit. per gram DOSE EQUIVALEst? I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or eaceeding the limit line shows en Figure 3.4-1, he la at least teor STApener with T, less than 500*r within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, b.

With the speettis activity of the primary emelant p l

them let/I missecurie per gram, he la at least Est StsmST teith T less than SOS *r withia 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

• With T greater than or equal to 500*r.

)

rAALEY-UlrIT 1 3/4 4-23 AMEN! STENT 30.44,446, 117 9704020296 970325 PDR ADOCK OS000348 P

PDR

- -.. - - - - -. -. -... - -. ~ -.. -. -... - - -... _. -

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REAC*0R COOLANT SYSTDI i

ACTION:

(ContinueO MODES 1, 2, 3, 4, and 5:

g.

~

a.

With th pecific activity of the primary coolant greater i

than WereCurie per gram DOSE EQUIVALENT I-131 or j

greater than 100 / E microcuries per gram, perform the sasyling a>4 analysis requirements of item (a of Table 4.4-4 until the specific activity of the primary coolent is restored to withis its limits.

SURVEILIANCE REQUIRDerTS 4.4.9 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

hl yAALEY-UNIT 1 3/4 4-24 AMENmert 30. W.117

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TYPs OF N 3s3333 m mis pegges 33 giggen sampgg n

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Greco Activ.ty Deteemination At least sees pee 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 1, 2, 3, 4 2.

Isotopte Ametyete for Dost 1 per 14 days 1

80UIVALEuff I-131 Conce.tratten 3.

Redtechendeel for i 1 Per 4, m

• 1 wg Determinettee r

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Isotopic Ametrete for Iodine el case per 4 homes, 18,28,30,48, SS l

w Includtag I-131, I-133, and I-135 eenesesse the spesada l

eettwity esecede 3

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I-131 or 199/8 pC1/ gram, and b) ese semple h 2 and 4 heute felleuteg a 1, 2, 3 m M cheogo eseeedtog It poseent et the m M Peuem

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PDCINF W MfS 15E300E, ptBElR FIEWS 3.4-1 k

1 DOSE W311TVALENT I-131 Primary Coelant Specific Activity Limit versus Percent of RATED TuBDSLL PCEER with the Priesty Caelaat Specific f

Activity > +rf1C1/ gram Dose Equivalent 3-131

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FAALEY-UNIT 1

,y 3/4 4-26 AMDtIMDif DC.84rH6,117 m.

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stActos cooturt systtM Basts 3/4.4.0 CMEMISTRY 1,,e umaeuens.

acte, ca.lant

,ste..eaist,y ensure met.e,,esies e, ae.eacter coolant Sy.t. is.ni..ed au,e.u.es me,etent,a1 ier a.te, l

Coelant System leakage or failure due to stress cortesiaa. Maantaanang sh l

chemistry within the Steady State Limits provides adeguate corresten protection to ensure the structural integrity of the.eacter coolant system ever the life of the plant. The associated effects of escoeding the saygen, chloride, and fluoride limits are time and temperature dependent. Corressoa studies show that operaties any be esatinued with contaminant eencontration levels in escess of the ' Steady State Limits, up to the Transient Lamats, for the specified limited time intervals wathout having a significant ef fect en the structural integrity of the peacter Coolant system. The time interval permitting eentinued operaties withis the restrictions of the Transtent Lamats provides time for taking sorteettwo actions to restore the contaminant canoestrations to withis the Steady State Limits.

The serwe111ames requiremmate provide adeguate assuranee that eencontrations is enenes of the limits utta he detected La suffisiest time to take sorteettwo setten.

3/4.4.9 SFectrIC ACT!vffY The limitettene en the apositis activity of the primary eselaat ensure that the resulting S hear desee at tia site boundary will met escoed an appropriately smali tsasties of Part 100 timits la the event of primary-te-l eeeendary leakage as a result of a steenline break.

The Act!ON statement pesudtting POWER OPERATION to contiamo for limited time periods with the primary ecolast's specific activity greater than 01 l

microcuries/ gram 8085 BOUIVA& art I-131, but withis the allowable limit been en Figure 3.4-1, aseemandates poesihte iodias spiking phemesmomen selsk may occur fe11 ewing changes la TussgebL POWER.

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Revised Technical Specification Pages J

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Pase 3/4 4-24 Replace Page 3/4 4-25 Replace Pase 3/4 4-26 Replace a

Page B 3/4 4-5 Replace l

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l REACTOR COOLANT SYSTEM i

3/4.4.9 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.9 The specific activity of the primary coolant shall be limited tot Less than or equal to 0.3 microcurie per gram DOSE l

a.

EQUIVALENT I-131; Less than or equal to 100/5 microcurie per gram.

b.

1 APPLICABILITY:

MODES 1, 2, 3, 4,

and 5 t

1 ACTION:

MODES 1, 2, and 3*:

With the specitic activity of the primary coolant greater j

a.

than 0.3 microcurie per gram DOSE EQUIVALENT I-131 for more l

than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with Tavg less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

I b.

With the specific activity of the primary coolant greater than 100/E microcurie por gram, be in at least HOT STANDBY with Tavg less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

1

• With Tavg greater than or equal to 500*F.

l FARLEY-UNIT 1 3/4 4-23 AMENDKENT NO.

I

a REACTOR ~ COOLANT SYSTEM ACTION:

(Continued)

NODES 1, 2, 3, 4, and 5:

With the specific activity of the primary coolant greater a.

than 0.3 microcurie per gram DOSE EQUIVALENT I-131 or l

greater than 10 0 / E microcuries per gram, perform the sampling and analysis requirements of item 4a of Table 4.4-4 until the specific activity of the primary coolant is restored to within its 1Laits.

SURVEILLANCE REQUIREMENTS 4.4.9 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

1 FARLEY-UNIT 1 3/4 4-24 AMENDMENT NO.

l l

)

o TABLE 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM 5

E TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN WHICH SAMPLE Q

AND ANALYSIS FREQUENCY AND ANALYSIS REOUIRED a

1.

Gross Activity Determination At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 1, 2, 3, 4 4

s 2.

Isotopic Analysis for DOSE 1 per 14 days 1

EQUIVALENT I-131 Concentration 1 Per 6 months

• 1 3.

Radiochemical for E Determination 4.

Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 1#, 2#, 3#, 4#, 5#

s Including I-131, I-133, and I-135 whenever the specific

?

activity exceeds 0.3 g

pC1/ gram DOSE EQUIVALENT I-131 or 100/E pCi/ gram, and b) One sample between 2 and 5

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a 1, 2, 3 6

THERMAL POWER change el exceeding 15 percent of the RATED THERMAL POWER within a one hour period.

.U

1. Until the specific activity of the primary coolant system is restored within its limits.

Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

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20 30 40 50 60 70 80 90 100 PERCENT OF PATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 0.3 pCi/ gram Dose Equivalent I-131 l

FARLEY-UNIT 1 3/4 4-26 AMENDMENT NO.

1

,e REACTOR COOLANT SYSTEM BASES 3/4.4.8 CHEMISTRY i

The limitations on Reactor Coolant System chemistry ensure that corrosion i

of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion.

I Maintaining the chemistry within the steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady i

State Lindts, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

1 The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3/4.4.9 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an i

appropriately small fraction of Part 100 limits in the event of primary-to-secondary leakage as a result of a steamline break.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.3 microcuries/ gram DOSE EQUIVALENT I-131, but within the allowable l

limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

1 l

i FARLEY-UNIT 1 B 3/4 4-5 AMENDMENT NO.

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i Enclosurc 4 i

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i Revised Responses to Generic Letter 95-05 Guidance i

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1 Revised Responses to Generie Letter 95-05 Guidance Southern Nuclear Operating Company (SNC) will implement the requested actions of Generic Letter 95-05 with the following comments:

(1)

The applicability requirements discussed in Section 1 of Attachment I of Generic Letter 95-05 will be splemented.

1.b.! - Concerning the deformation or collapse of steam generator tubes following a loss of coolant accident plus a safe shutdown carthquake event, a Farley specific analysis was j

docketed under WCAP 12871, Revision 2 dated February 1992. As a result of this analysis, no tubes will be excluded from using the voltage repair criteria.

(2)

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'Ihe inspection criteria discussed in Section 3 of Attachment I of the Generic Letter will be bp'ested with the followmg responses / clarifications. In addition, the inspection guidance will be implemented in accordance with the Appendix A guidelines last submitted to the NRC by letter dated February 23,1994.

3.b -

SNC will utilize a motorized rotating coil probe, e.g., pancake or + Point, instead of specifying a rotating pancake coil. This wording change is made to ensure that the + Point probe can be used as an alternative to the rotating pancake coil.

3.b.1 - SNC will inspect all bobbin flaw indications with voltages greater than 2.0 volts with a motorized rotating coil probe.

3.b.2 - SNC will inspect all intersections where copper signals interfere with the detection of flaws with a motorized rotating coil probe. Any indication found with the motorized rotating coil will result in repair of the tube.

i i

I 3.b.3 - All intersectons with dent signals greater than 5 volts will be inspected with a motorized rotating coil probe. Any indications found at such intersections with the motorized i

rotating coil probe will result in repair of the tube. If circumferential cracking or prhary j

water stress corrosion cracking indications are detected, the motorized rotating coil probe sampling plan may be expanded to include dents less than 5 volts.

SNC. "" :=pe : :!! i;;asectic= nii in: :i;=!: gr=:= in 5 vc!= wii : =c:=i:d rc:::!=g eci! preb. !f eircumfax:!:! = ki=g = p.-h=y =:==== :==i= =ching i: o?r'd -: S ::S ::pper, p'" : in:== :!c.:, ;==p'! g p!= vi'! k i ap'==::d in

=:=i = wii i: P"/R S:=. C:==:= Tub E=.!=:!= 0;i&H=:, R:.i;i= 4. If i

l indica:ic= : feud : &ne..ii !::;= ::= 5 vc!u, i: thu wi!! b eb=c:si=d. If j

i f.:.; =: i S==::=:! =;;;...=: cf R:g;!::=y "W42 !, i:== phag p'n 9" b =;=id c i;t-+M-.: nii d== kn S.- 5 vc!=. !f Se f': i: ev:!r"-f =

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,e-3.b.4 - SNC will inspect all intersections with large mixed residuals that could be expected to j

mask a 1.0 volt bobbin flaw signal with a motorized rotating coil probe.

3.c.2

'Ihe *10% limit on new probe variability will be implemented using the guidance included l

in Nuclear Energy Institute to NRC letter dated January 23,1996, concerning "New Probe Variability for Use in the ODSCC Alternate Repair Criteria", as discussed in the NRC to the Nuclear Energy lastitute letter dated February 9,19%. Furthermore, SNC will verify -

'i that both the primary and mix frequencies will meet the i10% variability requirement.

l 3.c.3 - The limits on probe wear will be implemented using the guidance included in Nuclear Energy to NRC letter dated January 23,1996, conceming " Eddy Current Probe Replacement Criteria for Use in ODSCC Alternate Repair Criteria"; as discussed in the NRC to se Nuclear Energy Institute letter dated February 9,1996; NEl to NRC letter-f dated February 23,1996; and NRC to NEI letter dated March 18,1996. The followmg l

summanzes this guidance as agreed to by the NRC Staff:

For all tubes identified with indications above 1.5 volts (i.e.,75% of the 2 vo:t repair limit for 7/8 inch tubes is 1.5 volts) since the last successful probe wear cher.k (< 15% wear), the whole tube (i.e., all hot-leg tube suppon plate interse.ctions to the lowest cold-leg TSP intersection with known ODSCC) will be re-inspected with an acceptable probe (<l5% probe wear) and all eddy current j

data from the acceatable probe will be evaluated. If a large indication (greater than approxunately I volt for 7/8 inch tubes) is A+c'ed which was previously missed with the failed probe, an assessment of the significance will be performed during the outage. This assessment, along with the description of actions taken,

)

will be provided to the NRC in the 90-day report.

The inspection described above will be modified slightly for tubes which would require a double entry to inspect the entire tube. For low row tubes in which the U-bend radius precludes passing a full size bobbin coil over the U-bend or for tubes with sleeves which preclude passing a full size bobbin through the sleeve, the portion of the tube with the indication above 75% of the repair limit I

will be re-inspected. The second entry for inspection of the remainder of the tube is not required provided there is not an indication above 75% of the repair limit.

Actions will be taken to minimize the potential for tubes to be inspected with probes that fail the probe wear check.1his includes replacing a probe unmediately upon fmdmg that it fails the probe wear check.

If a probe fails prior to performing a probe wear check, it will be assumed that the probe failed the probe wear check and the probe wear criteria approved by 1

the Staff will be followed.

The effects of probe wear will be explicitly assessed as a potential contributing factor if significant differences between the actual and end-of-cycle j

projections exist in the 90-day report.

i The 90-day report will address if a non-proportionate number of new indications have been detected in tubes which were inspected in the previous outage with a probe that failed the probe wear check.

3.c.4 - Data analysts will be tramed and qualified in the use of the analyst's guidelines and procedures. At Farley Nuclear Plant, a minimal number of analyss are used for i

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determination of voltage. The use of a small number of analysts is intended to minimize the effect of analyst variability on determination of growth rate, resulting in as accurate a prediction for the next operating cycle as possible. We believe this results in a more accurate growth rate determination; however, it is time consuming and can result in difficulty in performing the calculations prior to returning the steam generators to senice.

3.c.5 - Quantitative noise criteria have historically been applied and will be incorporated in the Farley Nuclear Plant Data Acquisition procedures. This enables noise levels due to electrical noise, tube noise, calibration standard noise, etc., to be addressed at the initial point ofinspection which has minimized the need for re-inspection. Probes are typically replaced prior to exceeding the noise criteria. If, upon measurement, the probe in use fails to meet the criteria, tubes tested with that probe since the last satisfactory measurement are re-inspected. In addition, the Farley Nuclear Plant Analysis procedures allow the analyst 1

to require re-inspection due to noise on a " qualitative" basis.

3.c.6 Data analysts will review the mixed residuals on the standard itself and take actions as necessary to minimize these residuals.

3.c.8 - Data analysts will be trained on the potential for prunary water stress corrosion cracking to occur at tube support plate intersections. The discovery of PWSCC at tube support plate intersections will be reportcJ t

.c NRC Staff prior to startup.

(2)

Calculations of the main steam line break leakage will be per the guidance of Sections 2.b and 2.c of Attachment 1 of Generic Letter 95-05 with the following responses:

2.b-Calculations performed in support of the voltage-based repair criteria will follow the probabilistic methodology described in WCAP-14277, Revision 1, SLB Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP Intersections. January M4 2.b.2(1) - No distribution cutoff will be applied to the voltage measurement variability distribution.

2.b.4 - In order to preclude the possible need for rapid turn around of a technical specification amendment for reactor coolant system specific iodine activity, the Farley Unit I technical specification will be reduced to 0.3 remaea@-> pCi/ gram. A leakage limit for Farley Unit I and4 of M 19 gpm is j=tif:ed in Att=h.= '. acceptable.

2.c-Reference is made to the use of an RPC probe. SNC will utilize a motorized rotating coil probe, e g., pancake or + Point, instead of specifying a rotating pancake coil. It is SNC's intent (and desire) to always perform the calculations on the projected EOC distributions.

In the event that the growth rate detenninations cannot be completed prior to returning the steam generators to senice, the calculations will be based on the actual EOC distributions as allowed in Section 2.c. However, even if the calculation made prior to returning the steam generators to senice is based on the actual measured voltage distribution, the l

l calculation based on the projected EOC voltage distribution will be provided to the NRC in l

the 90 day report following the outage.

4 (3)

Calculation of the conditional burst probability will be per the guidance of Section 2.a of Attachment I of Generic letter 95-05 with the following responses-2.a -

Calculations performed in support of the voltage-based repair criteria will follow the methodology described in WCAP-14277, Revision 1, SLB Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP Intersections., Jr - j 1995.

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Southern Nuclear will use the d*har forwarded to the NRC Staff by Duquesne Light Company letter dated March 27,1996, for the upcoming Farley Unit 1 inspection / evaluation. This is the same database that was used on the Farley Unit 2 inspection / evaluation. Southern Nuclear also requests that the NRC Staffprovide a projected schedule of review of the NEI database to allow utilities to make plans for upconung outages.

(4)

Farley leakage monitoring measures provide guidance on trending and response to rapidly increasing leaks. Guidance is provided not only for the absolute leakage measured, but also on the rate of change of the leak rate Timely Wion ofleaks is ensured by the N-16 monitors on both units. Farley has also implemented the guidelines contained in EPRI topical report "PWR Primary-to-Secondary Leak Guidelines," EPRI TR-104788, May 1995.-

(5)

Tube pull guidance of Section 4 of Attachment 1 of Generic Letter 95-05 will be followed.

(6)

Results will be reported per the guidance of Section 6 of Attachment I of Generic Letter 95-05.

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